ML24215A178
Text
Response to SDAA Audit Question Question Number: A-16-10 Receipt Date: 12/18/2023 Question:
Staff observes that the NuSCale US460 GTS Bases use, in most cases, assure, assures, and assured when ensure, ensures, and ensured, respectively, are more appropriate, and which conform to Rev. 5 of NUREG-1431 Bases. The applicant is requested to make conforming changes to the Bases.
Response
NuScale revises the GTS Bases and Technical Specification Sections 4.3 and 5.5.11 to replace assure, assures, and assured with ensure, ensures, and ensured, respectively.
Markups of the affected changes, as described in the response, are provided below:
NuScale Nonproprietary NuScale Nonproprietary
Design Features 4.0 NuScale US460 4.0-2 Draft Revision 2 4.0 DESIGN FEATURES 4.3 Fuel Storage (continued) b.
keff 0.95 if fully flooded with borated water at a minimum soluble boron concentration of [800] ppm, which includes an allowance for uncertainties to assureensure a 95 percent probability and 95 percent confidence level; c.
keff < 1.00 if fully flooded with unborated water, which includes an allowance for uncertainties to assureensure a 95 percent probability and 95 percent confidence level; d.
A nominal [10.00] inch center-to-center distance between fuel assemblies placed in the spent fuel storage racks.
4.3.2 Drainage The spent fuel pool is designed and shall be maintained to prevent inadvertent draining of the pool below [19] ft above the spent fuel pool floor.
4.3.3 Capacity The spent fuel pool is designed and shall be maintained with a storage capacity limited to no more than [600] fuel assemblies.
Programs and Manuals 5.5 NuScale US460 5.5-12 Draft Revision 2 5.5 Programs and Manuals 5.5.11 Surveillance Frequency Control Program This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assureensure the associated Limiting Conditions for Operation are met.
- a.
The Surveillance Frequency Control Program shall contain a list of Frequencies of those Surveillance Requirements for which the Frequency is controlled by the program.
- b.
Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, Risk-Informed Method for Control of Surveillance Frequencies, Revision 1. FSAR Table 16.1-1, Surveillance Frequency Control Program Base Frequencies, describes the plant licensing bases for the surveillance test intervals.
- c.
The provisions of Surveillance Requirements 3.0.2 and 3.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program.
[ 5.5.12 Spent Fuel Storage Rack Neutron Absorber Monitoring Program This Program provides controls for monitoring the condition of the neutron absorber used in the spent fuel pool storage racks to verify the Boron-10 areal density is consistent with the assumptions in the spent fuel pool criticality analysis. The program shall be in accordance with NEI 16-03-A, Guidance for Monitoring of Fixed Neutron Absorbers in Spent Fuel Pools, Revision 0, May 2017.]
SDM B 3.1.1 NuScale US460 B 3.1.1-1 Draft Revision 2 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.1 SHUTDOWN MARGIN (SDM)
BASES BACKGROUND According to GDC 26 (Ref. 1) the reactivity control systems must be redundant and capable of holding the reactor core subcritical when shutdown under cold conditions. Maintenance of the SDM ensures that postulated reactivity events will not damage the fuel.
SDM requirements provide sufficient reactivity margin to assureensure that specified acceptable fuel design limits (SAFDLs) will not be exceeded for normal shutdown and anticipated operational occurrences (AOOs). As such, the SDM defines the degree of subcriticality that would be obtained immediately following the insertion or scram of all shutdown and regulating bank control rod assemblies (CRAs), assuming that the single CRA of highest reactivity worth is fully withdrawn.
Additionally SDM requirements provide sufficient reactivity margin to ensure that the reactor will remain shutdown at all temperatures with all control rods inserted.
The system design requires that two independent reactivity control systems be provided, and that one of these systems be capable of maintaining the core subcritical under cold conditions. These requirements are provided by the use of movable CRAs and soluble boric acid in the Reactor Coolant System (RCS). The CRA System provides the SDM during power operation and is capable of making the core subcritical rapidly enough to prevent exceeding acceptable fuel damage limits, following all AOOs and postulated accidents, assuming that the CRA of highest reactivity worth remains withdrawn.
The soluble boron system can compensate for fuel depletion during operation and all xenon burnout reactivity changes and maintain the reactor subcritical under cold conditions.
During power operation, SDM control is ensured by operating with the shutdown bank groups fully withdrawn and the regulating bank groups within the limits of LCO 3.1.6, Regulating Bank Insertion Limits.
When the unit is in MODES 2, 3, 4 or 5, the SDM requirements are met by means of adjustments to the RCS boron concentration and the boron requirements for the pool, LCO 3.5.3, "Ultimate Heat Sink" and CRA controls.
Regulating Bank Insertion Limits B 3.1.6 NuScale US460 B 3.1.6-2 Draft Revision 2 BASES BACKGROUND (continued)
The power density at any point in the core must be limited so that the fuel design criteria are maintained. Together, LCO 3.1.4, Rod Group Alignment Limits, LCO 3.1.5, Shutdown Bank Insertion Limits, LCO 3.1.6, Regulating Bank Insertion Limits, LCO 3.2.1, Enthalpy Rise Hot Channel Factor (FH), and LCO 3.2.2, AXIAL OFFSET (AO) provide limits on control component operation and on monitored process variables which ensure that the core operates within the fuel design criteria.
The shutdown and regulating bank insertion and alignment limits and power distribution limits are process variables that together characterize and control the three dimensional power distribution of the reactor core. Additionally, the regulating bank insertion limits control the reactivity that could be added in the event of a rod ejection accident, and the shutdown and regulating bank insertion limits assureensure the required SDM is maintained.
Operation within the subject LCO limits will prevent fuel cladding failures that would breach the primary fission product barrier and release fission products to the reactor coolant in the event of a loss of coolant accident (LOCA), loss of flow, ejected CRA, or other accident requiring termination by a Reactor Trip System (RTS) trip function.
APPLICABLE SAFETY ANALYSES The regulating bank insertion limits, FH, and AO LCOs are required to prevent power distributions that could result in fuel cladding failures in the event of a LOCA, loss of flow, ejected CRA, or other accident requiring termination by an RTS trip function.
The acceptance criteria for addressing shutdown and regulating bank group insertion limits and inoperability or misalignment are that:
- a. With the most reactive CRA stuck out there will be no violations of either:
- 1. specified acceptable fuel design limits; or
- 2. Reactor Coolant System (RCS) pressure boundary integrity; and
- b. The core remains subcritical after design basis events with all CRAs fully inserted.
Rod Position Indication B 3.1.7 NuScale US460 B 3.1.7-2 Draft Revision 2 BASES BACKGROUND (continued)
The axial position of shutdown bank CRAs and regulating bank CRAs are determined by two separate and independent means: the Counter Position Indicators (CPIs) (commonly called bank step counters) and the Rod Position Indicators (RPIs).
The CPI monitors the commands sent to the CRDM coils from the Control Rod Drive System (CRDS) that moves the CRAs. There is one step counter for each CRDM. The CRA CPI is considered highly precise (+/- 1 step or +/- 0.375 inch). If a CRA does not move one step for each command signal, the step counter will still count the command and incorrectly reflect the position of the CRA.
The RPI function of the CRDS provides a highly accurate indication of actual CRA position, but at a lower precision than the step counters.
This system is based on inductive analog signals from a series of coils spaced along a hollow tube with a center to center distance of 1 inch.
To increase the reliability of the RPI system, the inductive coils of a CRA's two RPI channels are alternately connected to two separate data systems. Each RPI channel is associated with just one of the data systems. Thus, if one system fails, the RPI will go to reduced accuracy of 2.25 inches, which is 6 steps. The normal indication accuracy of the RPIs is +/- 0.5 inch, and the accuracy with one channel of RPI out-of-service is +/- 6 steps (+/- 2.25 inches).
APPLICABLE SAFETY ANALYSES The regulating and shutdown bank groups CRA position accuracy is essential during power operation. Power peaking, ejected CRA worth, or SDM limits may be violated in the event of a Design Basis Accident (Ref. 2), with regulating or shutdown bank CRAs operating outside their limits undetected. Therefore, the acceptance criteria for CRA position indication is that CRA positions must be known with sufficient accuracy in order to verify the core is operating within the group sequence, overlap, design peaking limits, ejected CRA worth, and within minimum SDM (LCO 3.1.5, Shutdown Bank Insertion Limits, LCO 3.1.6, Regulating Bank Insertion Limits). The CRA positions must also be known in order to verify the alignment limits are preserved (LCO 3.1.4, Rod Group Alignment Limits). CRA positions are continuously monitored to provide operators with information that assuresensures the unit is operating within the bounds of the accident analysis assumptions.
Boron Dilution Control B 3.1.9 NuScale US460 B 3.1.9-3 Draft Revision 2 BASES LCO The requirement that two demineralized water isolation valves be OPERABLE assuresensures that there will be redundant means available to terminate an inadvertent boron dilution event. The requirement that the boron concentration of the boric acid supply be maintained within the limits specified in the COLR ensures that the supply is not a source to the CVCS that could result in an inadvertent boron dilution event.
The limits on maximum CVCS makeup pump demineralized water flow path flowrate are established by restricting the flow that can be provided during system operation to within the limits in the COLR. The restrictions may be implemented by use of at least one closed manual or one closed and de-activated automatic valve, or by removing the power supply from one CVCS makeup pump.
APPLICABILITY The requirement that two demineralized water isolation valves be OPERABLE, and that the boric acid storage tank boron concentration and maximum CVCS makeup pump demineralized water flow path flowrate is within the limits specified in the COLR is applicable in MODES 1, 2, and 3 with any dilution source flow path in the CVCS makeup line not isolated. In these MODES, a boron dilution event is considered possible, and the automatic closure of these valves is assumed in the safety analysis. The boron concentration of the boric acid sources are not assumed to be capable of causing a dilution event by the boron dilution event analysis. The maximum CVCS makeup pump demineralized water flow path flowrate is an assumption of the boron dilution event.
In MODE 1 < 15% RTP, the detection and mitigation of a boron dilution event would be signaled by a High Source or Intermediate Range Log Power Rate or a High Source Range Count Rate.
In MODE 1 15% RTP, the detection and mitigation of a boron dilution event would be signaled by a High Power Range Rate or High Power Range Linear Power. In MODES 2 and 3, the detection and mitigation of a boron dilution event would be signaled by a Source Range High Count Rate trip, a trip on Source Range High Log Power Rate, or a trip on High Subcritical Multiplication, or low RCS flow.
In MODES 4 and 5, a dilution event is precluded because the CVCS RCS injection and discharge flow paths are not connected to the RCS, thus eliminating the possibility of a boron dilution event in the RCS.
Pool volume is sufficient to minimize the potential for boron dilution during MODE 5 within the surveillance intervals provided by LCO 3.5.3, Ultimate Heat Sink.
Boron Dilution Control B 3.1.9 NuScale US460 B 3.1.9-6 Draft Revision 2 BASES SURVEILLANCE REQUIREMENTS (continued)
Boron concentration in the supply is verified to be within the limits specified in the COLR by periodic measurement.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.1.9.4 This Surveillance verifies that CVCS makeup pump maximum flowrate is 25 gpm. The lowest maximum makeup pump demineralized water flowrate that can be used while in operation is that of one CVCS makeup pump as assumed in the boron dilution analysis. The Surveillance verifies the maximum flowrate of each CVCS makeup pump is consistent with the analysis assumptions. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.The limits on maximum CVCS makeup pump demineralized water flow path flowrate are established by restricting the flow that can be provided during system operation to within the limits in the COLR.
The restrictions may be implemented by use of at least one closed manual or one closed and de-activated automatic valve, or by removing the power supply from one CVCS makeup pump.
SR 3.1.9.5 This Surveillance verifies that the CVCS injection flow path is isolated from other module CVCS flow paths when the module heatup system (MHS) is in use. This ensures that dilution or other unplanned changes to boration cannot occur due to cross-flow between the module support systems. The module CVCS to MHS flow paths include dual isolation valves and instrumentation used to assureensure that isolation exists.
Two Notes modify SR 3.1.9.5. The first Note states that the SR is only required to be met when CVCS flow is aligned through the MHS. This is considered acceptable since cross-flow can only occur when the CVCS flow is aligned through the MHS. The second Note allows isolation devices that are locked, sealed, or otherwise secured in position to be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since the function of locking, sealing, or securing components is to ensure that these devices are not inadvertently repositioned. Therefore, the probability of misalignment of these devices once they have been verified to be in the proper position is small.
FH B 3.2.1 NuScale US460 B 3.2.1-4 Draft Revision 2 BASES SURVEILLANCE SR 3.2.1.1 REQUIREMENTS The value of FH is determined by using the fixed in-core instrument system to obtain a flux distribution map. A data reduction computer program then calculates the maximum value of FH from the measured flux distributions. The in-core instrument design and procedures incorporate the methods and process for measuring FH using the available in-core instrumentation. The procedures include verification that adequate instrument indications are available to provide a representative value of FH consistent with the methodology used to establish the FH limits in the COLR. This assuresensures that the FH is within limits of the LCO. After each refueling, FH must be determined in MODE 1 prior to exceeding 20% RTP. This requirement ensures that FH limits are met at the beginning of each fuel cycle and in accordance with the misload event analysis. (Ref. 1)
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES
- 1. FSAR, Chapter 15.
AO B 3.2.2 NuScale US460 B 3.2.2-3 Draft Revision 2 BASES SURVEILLANCE SR 3.2.2.1 REQUIREMENTS This Surveillance verifies that the AO, as indicated by the in-core instrumentation system, is within its specified limits.
The in-core instrument design and procedures incorporate the methods and process for verifying the AO is within limits using the available in-core instrumentation. The surveillance procedures include verification that adequate instrument indications are available to provide a representative value of the AO consistent with the methodology used to establish the AO limits in the COLR. This assuresensures that the AO is within limits of the LCO.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES 1.
FSAR, Chapter 15.
2.
FSAR, Chapter 4.
MPS Instrumentation B 3.3.1 NuScale US460 B 3.3.1-53 Draft Revision 2 BASES ACTIONS (continued)
If the Required Actions associated with this Condition cannot be completed within the required Completion Time, the unit must be brought to a MODE or other specified condition where the Required Actions do not apply. This is accomplished by isolating the dilution source flow paths in the CVCS makeup line by use of at least one closed manual or one closed and de-activated automatic valve. The allowed Completion Time for H.1 of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is reasonable, based on operating experience, for reaching the required condition in an orderly manner.
I.1 and I.2 Condition I is entered when Condition C applies to Functions that result in a DHRS, SSI, or ECCS actuation, as listed in Table 3.3.1-1.
If the Required Actions associated with this Condition cannot be completed within the required Completion Time, the unit must be brought to a MODE or other specified condition where the Required Actions do not apply. This is accomplished by Required Actions I.1 and I.2.
I.1 places the unit in MODE 2 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. This action limits the time the unit may continue to operate with a limited or inoperable automatic channel.
I.2 requires the unit to be in MODE 3 and PASSIVELY COOLED within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> of entering the Condition. These conditions assureensure adequate passive decay heat transfer to the UHS and result in the unit being in a condition for which the LCO no longer applies.
Completion Times are established considering the likelihood of a LOCA event that would require ECCS or DHRS actuation. They also provide adequate time to permit evaluation of conditions and restoration of channel OPERABILITY without challenging plant systems during a shutdown.
J.1 As listed in Table 3.3.1-1, Condition J is entered when Condition C applies to Function 27.a, "High RCS Pressure - Low Temperature Overpressure Protection (LTOP)," which results in actuation of the LTOP system.
MPS Instrumentation B 3.3.1 NuScale US460 B 3.3.1-54 Draft Revision 2 BASES ACTIONS (continued)
If a Required Action associated with Condition A or B cannot be completed within the required Completion Time, or three or more channels of this Function are inoperable, the unit must be brought to a MODE or other specified condition where the LCO and Required Actions for this Function do not apply. This is accomplished by opening at least one RVV. The Completion Time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is reasonable, based on operating experience, for establishing an RCS vent flow path sufficient to ensure low temperature overpressure protection.
K.1 and K.2 Condition K is entered when Condition C applies to Functions that result in actuation of the Containment Isolation system as listed in Table 3.3.1-1.
If the Required Actions associated with this Condition cannot be completed within the required Completion Time, the unit must be brought to a MODE in which the LCO does not apply. This is accomplished by Required Actions K.1 and K.2. K.1 places the unit in MODE 2 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. This action limits the time the unit may continue to operate with a limited or inoperable CIS automatic channel. K.2 places the unit in MODE 3 with RCS hot temperature < 200 °F within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of entering the Condition. This Condition assuresensures the unit will maintain the RCS depressurized and the unit being in a condition for which the LCO no longer applies.
Completion Times are established considering the likelihood of a design basis event that would require CIS actuation during the period of inoperability. They also provide adequate time to permit evaluation of conditions and restoration of channel OPERABILITY without challenging plant systems during a shutdown.
L.1, L.2, L.3, L.4, and L.5 Condition L is entered when Condition C applies to Functions that result in a reactor trip, CIS actuation, DHR actuation, DWSI, CVCSI, SSI, ECCS actuation, and Pressurizer Heater Trip due to the Low AC voltage to EDAS battery charger or High Under-the-Bioshield Temperature as listed in Table 3.3.1-1.
If the Required Actions associated with this Condition cannot be completed within the required Completion Time, the unit must be brought to a MODE or other specified condition where the Required Actions do not apply. This is accomplished by Required Actions L.1, L.2, L.3, L.4, and L.5.
MPS Instrumentation B 3.3.1 NuScale US460 B 3.3.1-55 Draft Revision 2 BASES ACTIONS (continued)
L.1 places the unit in MODE 2 within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. This action limits the time the unit may continue to operate with a limited or inoperable automatic channel. L.2 requires the unit to be in MODE 3 and PASSIVELY COOLED within 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> of entering the Condition. These conditions assureensure adequate passive decay heat transfer to the UHS and result in the unit being in a condition for which the DHRS OPERABILITY is no longer required.
L.3 places the unit in MODE 3 with RCS temperature below the T-2 interlock within 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> of entering the Condition. This Condition assuresensures the unit will maintain the RCS depressurized and the unit being in a condition for which the LCO no longer applies.
L.4 isolates the dilution source flow paths in the CVCS makeup line by use of at least one closed manual or one closed and de-activated automatic valve within 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />. This completes the function of the DWSI.
L.5 opens the power supply breakers to the pressurizer heaters within 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />.
Completion Times are established considering the likelihood of a design basis event that would require automatic actuation during the period of inoperability. They also provide adequate time to permit evaluation of conditions and restoration of channel OPERABILITY without challenging plant systems during a shutdown.
M.1 and M.2 Condition M is entered when Condition C applies to the following Functions as listed in Table 3.3.1-1.
8.d, Low Pressurizer Pressure - Pressurizer Line Isolation 22.b, High Narrow Range Containment Pressure - Containment Isolation (CIS) 22.c, High Narrow Range Containment Pressure - Decay Heat Removal System (DHRS) Actuation 22.d, High Narrow Range Containment Pressure - Secondary System Isolation (SSI)
MPS Instrumentation B 3.3.1 NuScale US460 B 3.3.1-56 Draft Revision 2 BASES ACTIONS (continued)
If the Required Actions associated with this Condition cannot be completed within the required Completion Time, the unit must be brought to a MODE or other specified condition in which the LCO and Required Actions for this Function does not apply. This is accomplished by Required Actions M.1 and M.2. M.1 places the unit in MODE 2 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. This action limits the time the unit may continue to operate with a limited or inoperable CIS automatic channel. M.2 places the unit in MODE 3 with RCS hot temperature below the T-3 interlock within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of entering the Condition. This Condition assuresensures the unit will be in a condition for which the LCO no longer applies.
Completion Times are established considering the likelihood of a design basis event that would require CIS actuation during the period of inoperability. They also provide adequate time to permit evaluation of conditions and restoration of channel OPERABILITY without challenging plant systems during a shutdown.
N.1 Condition N is entered when Condition C applies to Functions that result in a reactor trip, CIS, SSI, DWSI, DHRS actuations when above the RCS narrow range T-4 interlock as listed in Table 3.3.1-1.
If the Required Actions associated with this Condition cannot be completed within the required Completion Time, the unit must be brought to a specified condition in which the LCO and Required Actions for this Function does not apply. This is accomplished by reducing the narrow range RCS temperature below the T-4 interlock within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. This Condition assuresensures the unit will be in a condition for which the LCO no longer applies.
The Completion Time was established considering the likelihood of a design basis event that would require the actuation during the period of inoperability. It also provides adequate time to permit evaluation of conditions and restoration of channel OPERABILITY without challenging plant systems to reach the required configuration.
MPS Instrumentation B 3.3.1 NuScale US460 B 3.3.1-60 Draft Revision 2 BASES SURVEILLANCE REQUIREMENTS (continued)
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.3.1.5 SR 3.3.1.5 is the performance of a CHANNEL CALIBRATION of the Class 1E isolation devices, as described in SR 3.3.1.4.
Class 1E isolation devices ensure that electrical power to the associated MPS circuitry and logic will not adversely affect the ability of the system to perform its safety functions. The devices de-energize and isolate the MPS components if such a condition is detected. This surveillance verifies the setpoints and functions of the isolation devices including associated alarms and indications by performing a CHANNEL CALIBRATION of required Class 1E isolation devices. The overcurrent and undervoltage setpoints of the Class 1E isolation devices are established and controlled in accordance with the Setpoint Program. The calibration parameters associated with the CHANNEL CALIBRATION of these Class 1E isolation devices are established to assureensure component OPERABILITY of the device electrical protection and isolation functions. There are no LSSSs associated with the Class 1E devices such that the establishment of a limiting trip setpoint (LTSP) or nominal trip setpoint (NTSP) is not governed by the Setpoint Program. However, the performance of a CHANNEL CALIBRATION implements sections of the Setpoint Program and includes the channel OPERABILITY determination based on the As-Found and As-Left settings for the Class 1E device calibration parameters.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES
- 1. Regulatory Guide 1.105, Revision 4, February 2021.
- 3. 10 CFR 50.34a.
- 4. FSAR, Chapter 7.
- 5. FSAR, Chapter 15.
- 6. 10 CFR 50.49.
RTS Logic and Actuation B 3.3.2 NuScale US460 B 3.3.2-5 Draft Revision 2 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.2.2 This SR measures the ACTUATION RESPONSE TIME of the RTS divisions. The ACTUATION RESPONSE TIME is combined with the allocated MPS digital time response and the CHANNEL RESPONSE TIME to determine and verify the TOTAL RESPONSE TIME is less than or equal to the maximum values assumed in the safety analysis.
Individual component response times are not modeled in the analyses.
The analyses model the overall or total elapsed time, from the point at which the process variable exceeds the trip setpoint value at the sensor to the time at which the RTBs open. TOTAL RESPONSE TIME may be verified by any series of sequential, overlapping, or total division measurements.
CHANNEL RESPONSE TIMES are tested in accordance with LCO 3.3.1.
The maximum digital time response is described in the FSAR. This SR encompasses the ACTUATION RESPONSE TIME of the RTS division from the output of the equipment interface modules until the RTBs are open.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.3.2.3 SR 3.3.2.3 is the performance of a CHANNEL CALIBRATION of the Class 1E isolation devices, as described in SR 3.3.1.4.
Class 1E isolation devices ensure that electrical power to the associated MPS circuitry and logic will not adversely affect the ability of the system to perform its safety function. The devices de-energize and isolate the MPS components if such a condition is detected. This surveillance verifies the setpoints and functions of the isolation devices including associated alarms and indications by performing a CHANNEL CALIBRATION of required Class 1E isolation devices.
The overcurrent and undervoltage setpoints of the Class 1E isolation devices are established and controlled in accordance with the Setpoint Program. The calibration parameters associated with the CHANNEL CALIBRATION of these Class 1E isolation devices are established to assureensure component OPERABILITY of the device electrical protection and isolation functions. There are no LSSSs associated with the Class 1E devices such that the establishment of a limiting trip setpoint (LTSP) or nominal trip setpoint (NTSP) is not governed by the Setpoint Program.
ESFAS Logic and Actuation B 3.3.3 NuScale US460 B 3.3.3-7 Draft Revision 2 BASES ACTIONS (continued)
C.2 requires the unit to be in MODE 3 and PASSIVELY COOLED within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> of entering the Condition. This Condition assuresensures adequate passive decay heat transfer to the UHS and result in the unit being in a condition which assuresensures passive cooling of the reactor core.
Completion Times are established considering the likelihood of a LOCA event that would require ECCS, DHRS, or SSI actuation. They also provide adequate time to permit evaluation of conditions and restoration of actuation logic OPERABILITY without challenging plant systems during a shutdown.
D.1 If Required Action B.1 directs entry into Condition D as specified in Table 3.3.3-1, or if both divisions of the containment isolation actuation Function are inoperable then the unit is outside its design basis ability to automatically mitigate some design basis events.
With one division of actuation logic inoperable, the redundant signal paths and logic of the OPERABLE division provide sufficient capability to automatically actuate the CIS if required.
D.1 requires the unit to be placed in MODE 3 with RCS temperature below the T-2 interlock within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of entering the Condition. This condition assuresensures the unit will maintain the RCS depressurized, and the unit being in a condition for which the LCO no longer applies.
Completion Times are established considering the low probability of a design basis event that would require CIS actuation during the period of inoperability. They also provide adequate time to permit evaluation of conditions and restoration of actuation logic OPERABILITY without challenging plant systems during a shutdown.
E.1 If Required Action B.1 directs entry into Condition E as specified in Table 3.3.3-1, or if both divisions of demineralized water supply isolation actuation are inoperable then the unit is outside its design basis ability to automatically mitigate some design basis events.
ESFAS Logic and Actuation B 3.3.3 NuScale US460 B 3.3.3-12 Draft Revision 2 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.3.4 SR 3.3.3.4 is the performance of a CHANNEL CALIBRATION of the Class 1E isolation devices, as described in SR 3.3.1.4.
Class 1E isolation devices ensure that electrical power to the associated MPS circuitry and logic will not adversely affect the ability of the system to perform its safety functions. The devices de-energize and isolate the MPS components if such a condition is detected. This surveillance verifies the setpoints and functions of the isolation devices including associated alarms and indications by performing a CHANNEL CALIBRATION of required Class 1E isolation devices. The overcurrent and undervoltage setpoints of the Class 1E isolation devices are established and controlled in accordance with the Setpoint Program. The calibration parameters associated with the CHANNEL CALIBRATION of these Class 1E isolation devices are established to assureensure component OPERABILITY of the device electrical protection and isolation functions. There are no LSSSs associated with the Class 1E devices such that the establishment of a limiting trip setpoint (LTSP) or nominal trip setpoint (NTSP) is not governed by the Setpoint Program. However, the performance of a CHANNEL CALIBRATION implements sections of the Setpoint Program and includes the channel OPERABILITY determination based on the As-Found and As-Left settings for the Class 1E device calibration parameters.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.3.3.5 SR 3.3.3.5 verifies the pressurizer heater breaker actuates to the open position on an actual or simulated trip signal on each pressurizer heater breaker. This test verifies OPERABILITY by actuation of the end devices.
The pressurizer heater breaker test verifies the under voltage trip mechanism opens the breaker.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES
- 1. FSAR, Chapter 7.
RCS Pressure, Temperature, and Flow Resistance CHF Limits B 3.4.1 NuScale US460 B 3.4.1-1 Draft Revision 2 B 3.4 REACTOR COOLANT SYSTEM (RCS)
B 3.4.1 RCS Pressure, Temperature, and Flow Resistance Critical Heat Flux (CHF) Limits BASES BACKGROUND These Bases address requirements for maintaining RCS pressure and temperature within the limits assumed in the safety analyses. The safety analyses (Ref. 1) of normal operating conditions and anticipated operational occurrences assume initial conditions within the normal steady state envelope of operating conditions. For a given RCS flow resistance, RCS pressure and temperature in combination with THERMAL POWER establish the flow through the RCS including the reactor core. The limits placed on RCS pressure and temperature, in combination with the reactor power, ensure that the minimum critical heat flux ratio (CHFR) will be met for each of the transients analyzed.
The RCS pressure limit is consistent with operation within the nominal operational envelope. Pressurizer pressure indications are used to determine a value for comparison to the limit. A pressure below the limit will cause the reactor core to approach CHFR limits.
The RCS coolant cold temperature limit is consistent with full power operation within the nominal operational envelope. Indications of cold coolant temperature are averaged to determine a value for comparison to the limit. An RCS cold temperature above the limit could cause the core to approach CHF limits.
RCS flow resistance above the limit could cause a reduction in RCS flow and cause the core to approach CHF limits. The RCS flow resistance limit is consistent with and assuresensures that the flow rates assumed in the safety analyses will occur.
Operation for significant periods of time outside these CHF limits increases the likelihood of a fuel cladding failure in a CHF limited event.
APPLICABLE The requirements of this LCO represent the initial conditions for CHF SAFETY limited transients analyzed in the plant safety analyses (Ref. 1). The ANALYSES safety analyses have shown transients initiated within the requirements of this LCO will result in meeting the CHFR criterion. This is the acceptance limit for the RCS CHF parameters. Changes to the unit which could impact these parameters must be assessed for their impact on the CHFR criterion.
CVCS Isolation Valves B 3.4.6 NuScale US460 B 3.4.6-2 Draft Revision 2 BASES LCO The requirement that two CVCS isolation valves be OPERABLE for each of the four flow path lines connected to the RCS assuresensures that there will be redundant means available to isolate the CVCS from the RCS during a non-LOCA event or a steam generator tube failure accident should that become necessary. Also, the OPERABLE CVCS isolation valves provide isolation protection against postulated breaks outside of containment.
APPLICABILITY The requirement that two CVCS isolation valves for each of the four flow path lines connected to the RCS be OPERABLE is applicable in MODES 1, 2, and 3 because a pressurizer overfill event, steam generator tube failure accident, CVCS postulated break outside containment is considered possible in these MODES, and the automatic closure of these valves is assumed in the safety analysis.
In the applicable MODES, the need to isolate the CVCS makeup to the RCS is detected by the pressurizer level instruments, pressurizer pressure instruments, or containment pressure.
This isolation function is not required in MODES 4 and 5. In these MODES, pressurizer overfill, steam generator overfill, CVCS breaks outside containment during startup is prevented by unit conditions.
ACTIONS The ACTIONS are modified by two notes. Note 1 allows isolated penetration flow paths to be unisolated intermittently under administrative controls. These administrative controls consist of stationing a dedicated operator at the device controls, who is in continuous communication with the control room. In this way, the penetration can be rapidly isolated when a need for containment isolation is indicated.
Note 2 provides clarification that, for this LCO, separate Condition entry is allowed for each penetration flow path. This is acceptable, since the Required Actions for each Condition provide appropriate compensatory actions for each inoperable containment isolation device. Complying with the Required Actions may allow for continued operation, and subsequent inoperable CVCS isolation valves are governed by subsequent Condition entry and application of associated Required Actions.
CVCS Isolation Valves B 3.4.6 NuScale US460 B 3.4.6-4 Draft Revision 2 BASES ACTIONS (continued)
B.1 With two CVCS isolation valves in one or more penetration flow paths inoperable, the affected penetration flow path must be isolated within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The method of isolation must include the use of at least one isolation device that cannot be adversely affected by a single active failure. Isolation devices that meet this criterion are a closed and deactivated automatic valve, a closed manual valve, and a blind flange.
The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time is consistent with the ACTIONS of LCO 3.6.2. In the event the affected penetration is isolated in accordance with Required Action B.1, the affected penetration must be verified to be isolated on a periodic basis per Required Action A.2, which remains in effect. This periodic verification is necessary to assureensure leak tightness of containment and that penetrations requiring isolation following an accident are isolated. The Completion Time of once per 31 days for verifying each affected penetration flow path is isolated is appropriate considering the fact that the devices are operated under administrative controls and the probability of the misalignment is low.
C.1 and C.2 If the Required Actions and associated Completion Times are not met, the unit must be brought to a MODE or condition in which containment isolation requirement no longer applies. To achieve this status, the unit must be brought to at least MODE 2 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 3 with RCS hot temperature < 200 °F within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. The allowed Completion Times are reasonable to reach the required unit conditions from full power conditions in an orderly manner.
SURVEILLANCE SR 3.4.6.1 REQUIREMENTS This SR [applies to valves with actuators that incorporate pressurized accumulators as a source of stored energy. The SR] verifies adequate pressure in the accumulators required for CVCS isolation valve OPERABILITY. The pressure limits required for OPERABILITY, including consideration of temperature effects on those limits, applicable to the valve accumulators are established and maintained in accordance with the INSERVICE TESTING PROGRAM. The Frequency is controlled under the Surveillance Frequency Control Program.
DHRS B 3.5.2 NuScale US460 B 3.5.2-5 Draft Revision 2 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.5.2.3 Verification that the level in a steam generator (SG) is > 5% and 50% when its associated feedwater isolation valve is closed and the DHR system is not actuated assuresensures that the SG contains inventory adequate to support actuation and OPERABILITY of the associated decay heat removal system loop if it is required.
A Note Is provided Indicating that the surveillance Is not required to be performed when the associated FWIV is open. In those conditions, the normal feedwater system controls ensure that the SG will support DHRS OPERABILITY if it is required.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.5.2.4 Verification that the DHRS actuation valves are OPERABLE by stroking the valves open ensures that each loop of DHRS will function as designed when these valves are actuated. The DHRS actuation valves safety function is to open as described in the safety analysis.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.5.2.5 Verifying that the open ACTUATION RESPONSE TIME of each DHRS actuation valve is within limits is required to demonstrate OPERABILITY. The ACTUATION RESPONSE TIME is combined with the allocated MPS digital time response and the CHANNEL RESPONSE TIME to determine and verify the TOTAL RESPONSE TIME is less than or equal to the maximum values assumed in the safety analysis. The opening times are as specified in the INSERVICE TESTING PROGRAM. Each loop of DHRS contains two actuation valves, one actuated from each division of the MPS ESFAS actuation logic.
ACTUATION RESPONSE TIME is measured from output of the module protection system equipment interface module until the valves are open.
Ultimate Heat Sink B 3.5.3 NuScale US460 B 3.5.3-2 Draft Revision 2 BASES BACKGROUND (continued) provided in FSAR Chapter 15 (Ref. 2).
During transients and shutdowns which are not associated with design basis events in which DHRS or ECCS is actuated, water from the RP is added to the containment vessel by the Containment Flooding and Drain System (CFDS). After reaching an appropriate level in the containment, the reactor vent valves (RVVs) and reactor recirculation valves (RRVs) are opened to permit improved heat transfer from the reactor coolant system (RCS) to the containment vessel walls.
During normal operations, the RP limits temperatures of the module because the containment vessel is partially immersed in water. The water also provides shielding above and around the region of the core during reactor operations, limiting exposure to personnel and equipment in the area.
In MODE 4, the module is transported from the operating position to the RFP area of the UHS.
APPLICABLE SAFETY ANALYSES During all MODES of operation and storage of irradiated fuel, the UHS supports multiple safety functions.
The UHS level is assumed and credited in a number of transient analyses. A UHS level of 52 ft provides margin above the minimum level required to support DHRS and ECCS operation in response to LOCA and non-LOCA design basis events. The 52 ft level also assuresensures the containment vessel wall temperature initial condition assumed in the peak containment pressure analysis. The upper limit of 54 ft for the maximum pool level is an initial condition that ensures long term cooling analyses assumptions.
The UHS bulk average temperature is assumed and credited, directly or indirectly in design basis accidents including those that require DHRS and ECCS operation such as LOCA and non-LOCA design basis events. The bulk average temperature is also assumed as an initial condition of the peak containment pressure analysis, and the minimum pool temperature is an assumption used in long-term cooling analyses.
Containment B 3.6.1 NuScale US460 B 3.6.1-2 Draft Revision 2 BASES BACKGROUND (continued)
- b. De-activated automatic valves secured in their closed positions, except as provided in LCO 3.6.2, Containment Isolation Valves; and
- c. The sealing mechanism associated with each containment penetration (e.g. welds, flanges, or o-rings) is OPERABLE (i.e.,
OPERABLE such that the containment leakage limits are met).
APPLICABLE The safety design basis for the containment is that the containment SAFETY must withstand the pressures and temperatures of the limiting Design ANALYSES Basis Accident (DBA) without exceeding the design leakage rates.
The DBAs that result in a challenge to containment OPERABILITY from high pressures and temperatures are a loss of coolant accident (LOCA), a steam line break, and a rod ejection accident (REA) (Ref. 2). In addition, release of significant fission product radioactivity within containment can occur from a LOCA or REA. The DBA analyses assume that the containment is OPERABLE such that, for the DBAs involving release of fission product radioactivity, release to the environment is controlled by the rate of containment leakage. The containment is designed with an allowable leakage rate of 0.20% per day of containment air weight after a DBA (Ref. 3). This leakage rate, used in the evaluation of offsite doses resulting from accidents, is defined in 10 CFR 50, Appendix J (Ref. 1), as La: the maximum allowable containment leakage rate at the calculated peak containment internal pressure [940] psia (Pa) resulting from the limiting DBA. The allowable leakage rate represented by La forms the basis for the acceptance criteria imposed on containment leakage rate testing. La is assumed to be 0.20% per day in the safety analysis.
Satisfactory leakage rate test results are a requirement for the establishment of containment OPERABILITY.
The containment satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).
LCO The containment is designed to maintain leakage integrity < 1.0 La.
Leakage integrity is assuredensured by performing local leak rate testing (LLRT) and containment inservice inspection. Total LLRT leakage is maintained < 0.60 La in accordance with 10 CFR 50, Appendix J (Ref. 1).
Satisfactory LLRT and ISI examination are required for containment OPERABILITY.
Containment Isolation Valves B 3.6.2 NuScale US460 B 3.6.2-5 Draft Revision 2 BASES ACTIONS (continued)
B.1 Condition B has been modified by a note indicating that this Condition is only applicable to those penetration flow paths with two containment isolation valves.
With two containment isolation valves in one or more penetration flow paths inoperable, the affected penetration flow path must be isolated within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The method of isolation must include the use of at least one isolation device that cannot be adversely affected by a single active failure. Isolation devices that meet this criterion are a closed and de-activated automatic valve, a closed manual valve, or a blind flange.
The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time is consistent with the ACTIONS of LCO 3.6.1. In the event the affected penetration is isolated in accordance with Required Action B.1, the affected penetration must be verified to be isolated on a periodic basis per Required Action A.2, which remains in effect. This periodic verification is necessary to assureensure leak tightness of containment and that penetrations requiring isolation following an accident are isolated. The Completion Time of once per 31 days for verifying each affected penetration flow path is isolated is appropriate considering the fact that the devices are operated under administrative controls and the probability of the misalignment is low.
C.1 and C.2 If the Required Actions and associated Completion Times are not met, the unit must be brought to a MODE or condition in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 2 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 3 with RCS hot temperature < 200 °F within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
Completion Times are established considering the likelihood of an event that would require CIS actuation. They also provide adequate time to reach the required unit condition from full power conditions in an orderly manner.
Containment Closure B 3.6.3 NuScale US460 B 3.6.3-2 Draft Revision 2 BASES BACKGROUND (continued)
The closed containment ensures retention of adequate inventory to ensure transport of decay heat from the reactor core to the containment and UHS. The containment also limits the postulated release of radioactive fission products that could be released from the reactor core and reactor vessel.
The requirements for containment penetration closure ensure a loss of coolant inventory from the module will be restricted and ensures adequate core cooling and heat transfer to the UHS.
Containment penetrations that provide direct access from containment atmosphere to outside atmosphere must be isolated on at least one side.
Isolation may be achieved by a closed automatic isolation valve, or by a manual isolation valve, blind flange, or equivalent.
Equivalent isolation methods must be approved and may include use of a material that can provide an appropriately designed temporary barrier for the containment penetrations during MODE 3 and PASSIVELY COOLED and MODE 4.
APPLICABLE Passive core cooling in MODE 3 and MODE 4 with the upper module SAFETY assembly seated on the lower containment vessel flange requires the ANALYSES module liquid inventory to maintain core coverage and transfer decay from the reactor fuel to the ultimate heat sink. Containment closure ensures the inventory will remain available to perform this function.
Containment inventory prior to MODE 3 with PASSIVE COOLING is assuredensured by requirements for system OPERABILITY in LCO 3.5.1, "Emergency Core Cooling System," LCO 3.5.2, and "Decay Heat Removal System," and the definition of PASSIVE COOLING.
Containment inventory prior to MODE 4 when placing the upper module assembly on the lower containment vessel flange is assuredensured by the pool level requirements of LCO 3.5.3, Ultimate Heat Sink. These requirements assureensure the assumed inventory exists when the Applicability is met.
The closed containment provides a passive boundary to maintain the required coolant inventory and is a part of a primary success path that functions to prevent a design basis accident or transient that may otherwise challenge the integrity of a fission product barrier.
Containment Closure B 3.6.3 NuScale US460 B 3.6.3-3 Draft Revision 2 BASES APPLICABLE SAFETY ANALYSES (Continued)
Requiring containment closure in MODE 3 with PASSIVE COOLING, and MODE 4 when the upper module assembly is seated on the lower containment vessel flange satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).
LCO
[----------------------------------REVIEWERS NOTE-----------------------------------
The allowance to have penetration flow paths with direct access from the containment atmosphere to the reactor building atmosphere to be unisolated in MODE 3 and PASSIVELY COOLED and MODE 4 is based on:
(1) One or more RVV and one or more RRV open and the containment water level greater than or equal to 45 ft; and (2) Commitments from the licensee to implement acceptable administrative procedures to ensure that if an event occurs that challenges retention of adequate module inventory, the open penetration(s) can and will be promptly closed.
The time to close such penetrations or combination of penetrations shall be specified in the procedures.
]
This LCO limits the consequences of an event that could result in the loss of module inventory by limiting the potential release paths. The LCO requires any penetration providing a path from the containment to the reactor building atmosphere to be closed.
This LCO is modified by a Note allowing penetration flow paths to be unisolated under administrative controls. Administrative controls ensure
- 1) appropriate personnel are aware of the open status of the penetration flow path, and 2) specified individuals are designated and readily available to isolate the flow path in the event of an event challenging the ability to maintain adequate inventory in the module.
APPLICABILITY The containment closure requirements are applicable during operation in MODE 3 and PASSIVELY COOLED and MODE 4 with the upper module assembly seated on lower containment vessel flange. In MODES 1, 2, and 3 not PASSIVELY COOLED, core cooling and decay heat transfer to the UHS is assuredensured by:
LCO 3.5.1, Emergency Core Cooling System; LCO 3.5.2, Decay Heat Removal System (DHRS);
Containment Closure B 3.6.3 NuScale US460 B 3.6.3-4 Draft Revision 2 BASES APPLICABILITY (continued)
LCO 3.5.3, Ultimate Heat Sink; LCO 3.6.1, Containment; LCO 3.6.2, Containment Isolation Valves, LCO 3.7.1, Main Steam Isolation Valves (MSIVs); and LCO 3.7.2, Feedwater Isolation; In MODE 4 with the upper module assembly not seated on the lower containment vessel flange, and in MODE 5, core cooling and decay heat transfer to the UHS is assuredensured by LCO 3.5.3, Ultimate Heat Sink.
ACTIONS A.1 If any containment penetration is not in the required status, the penetration must be closed or otherwise placed in a condition where closure of the penetration is not needed. This is accomplished by immediately initiating action to close the containment or by establishing administrative controls that will ensure closure if required. Performance of this action shall not preclude completion of movement of a module to a safe position to accomplish closure.
SURVEILLANCE SR 3.6.3.1 REQUIREMENTS This Surveillance demonstrates that the penetrations required to be closed are in that position. This ensures module inventory will remain available to provide core cooling and transfer heat to the UHS via the containment vessel.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES
- 1. FSAR, Section 6.2.
MSIVs B 3.7.1 NuScale US460 B 3.7.1-3 Draft Revision 2 BASES APPLICABILITY (continued) through the valve isolated. When these valves are closed or their flow path is isolated, the required function has been satisfied. In MODE 3 when PASSIVELY COOLED, and in MODES 4 and 5, the unit is shutdown, the SGs do not contain significant energy or inventory, and the valves do not perform any credited safety function.
ACTIONS The ACTIONS are modified by a Note indicating that steam line flow paths may be unisolated intermittently under administrative control. These administrative controls consist of stationing a dedicated operator at the device controls, who is in continuous communication with the control room. In this way, the MSIV flow path can be rapidly isolated when a need is indicated.
A.1 and A.2 Condition A is modified by a Note stating that a separate Condition entry is allowed for each valve. This is acceptable because the Required Actions provide appropriate compensatory actions for each inoperable isolation valve. The series-parallel valve arrangement could result in multiple valves being inoperable and the redundant capability to isolate the steam line maintained.
With a required valve open and inoperable, isolation of the main steam flow using that valve to perform the credited isolation function can no longer be assuredensured. The isolation function could be susceptible to a single failure because only the redundant isolation valves on the affected steam line maintain the ability to isolate the effected steam flow.
Action A.1 requires isolation of the inoperable valve flow path within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Some repairs may be accomplished within the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> period to restore OPERABILITY and exit the Condition. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is reasonable because the inoperable isolation valve only affects the capability of one of the two redundant isolation valves to function. Only if a single failure occurs that affects the remaining capability to isolate the steam flow path will the safety function be affected.
The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is reasonable considering the availability of other means of mitigating design basis events, including Emergency Core Cooling System and the low probability of an accident occurring during this time period that would require closure of the specific flow path.
Nuclear Instrumentation B 3.8.1 NuScale US460 B 3.8.1-1 Draft Revision 2 B 3.8 REFUELING OPERATIONS B 3.8.1 Nuclear Instrumentation BASES BACKGROUND Three refueling neutron flux channels are provided to monitor the core reactivity during refueling operations. These detectors are located external to the reactor vessel below the reactor vessel flange and detect neutrons leaking from the core with the ability to be extended and retracted to facilitate module disassembly and reassembly.
The refueling neutron flux detectors are proportional counters. The detectors monitor the neutron flux in counts per second. The instrument range covers four decades of neutron flux (from 1E-1 cps to 1E3 cps) with approximately 5% instrument accuracy. The refueling neutron flux channels also provide continuous visual indication in the control room and continuous visual and audible indication at the refueling panel located in the reactor building in close proximity to the refueling area.
After the RPV is placed on the RPV refueling stand, a retractable support mechanism positions the refuel neutron monitors in close proximity to the RPV. This ensures the refuel neutron monitors are placed in the same position for each refueling. The refuel neutron monitors are located in the refuel pool bay area and are separate from the normal excore detectors used during operation. These are the only neutron monitors utilized during refueling.
APPLICABLE Two OPERABLE refueling neutron flux channels are required to SAFETY provide a signal to alert the operator to unexpected changes in core ANALYSES reactivity. During initial fuel loading, or when otherwise required, temporary neutron detectors may be used to provide additional reactivity monitoring (Ref. 1).
The audible count rate from the refueling neutron flux channels provides prompt and definite indication of any change in reactivity. The count rate increase is proportional to subcritical multiplication and allows operators to promptly recognize any change in reactivity. Prompt recognition of unintended reactivity changes is consistent with the assumptions of the safety analysis and is necessary to assureensure sufficient time is available to initiate action before SHUTDOWN MARGIN is lost (Ref. 2).
The refueling neutron flux channels satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).
Nuclear Instrumentation B 3.8.1 NuScale US460 B 3.8.1-2 Draft Revision 2 BASES LCO This LCO requires two of the three refueling neutron flux channels to be OPERABLE to ensure that redundant monitoring capability is available to detect changes in core reactivity during removal of the upper reactor vessel assembly and during fuel movement in the reactor vessel. To be OPERABLE, each channel must provide visual indication in the control room. In addition, at least one of the two required channels must provide an OPERABLE audible count rate function to alert the operators to the initiation of a boron dilution event.
APPLICABILITY In MODE 5 when the reactor vessel upper assembly is not seated on the reactor vessel flange, the refueling neutron flux channels are required to be OPERABLE to determine possible unexpected changes in core reactivity. There are no other direct means available to monitor the core reactivity conditions. The Applicability allows the retractable refueling neutron flux channels to be installed on the lower reactor vessel assembly following entry into MODE 5 (i.e., after detensioning the first reactor vessel flange bolt) and prior to the reactor vessel upper assembly lift. In MODES 1, 2, and 3 the Module Protection System neutron detectors and associated circuitry are required to be OPERABLE by LCO 3.3.1, Module Protection System (MPS) Instrumentation. In MODE 4, the module is disconnected from unborated water sources and the module Neutron Monitoring System. No changes to the core reactivity can occur in MODE 4 because a boron dilution event or fuel loading error cannot occur in this condition. Therefore, neutron monitoring is not required in MODE 4.
ACTIONS A.1 and A.2 Redundancy has been lost if only one refueling neutron flux channel is OPERABLE. In addition, if the required refueling neutron flux audible count rate channel is inoperable, prompt and definite indication of a boron dilution event, consistent with the assumptions of the safety analysis, is lost. Since these instruments are the only direct means of monitoring core reactivity conditions, positive reactivity additions, and introduction of water into the ultimate heat sink (UHS) with boron concentration less than required to meet the minimum boron concentration of LCO 3.5.3, Ultimate Heat Sink, must be suspended immediately. Suspending positive reactivity additions that could result in failure to meet the minimum boron concentration limit is required to assureensure continued safe operation. Introduction of water inventory must be from sources that have a boron concentration greater than that which would be required in the UHS for minimum refueling boron concentration. This may result in an