ML24348A062
| ML24348A062 | |
| Person / Time | |
|---|---|
| Site: | 05200050 |
| Issue date: | 12/13/2024 |
| From: | NuScale |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML24348A006 | List:
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| References | |
| LO-176318 | |
| Download: ML24348A062 (1) | |
Text
Response to NuScale Topical Report Audit Question Question Number: A-NonLOCA.LTR-43 Receipt Date: 06/17/2024 Question:
Several sections of TR-0516-49416, Revision 4, include language regarding operator actions, the implementation of which has not clearly been consistent with Limitation & Condition 5 of TR-0516-49416-P-A, Revision 3, which states, An applicant or licensee seeking to apply this methodology to a design must receive a separate approval through that design review for the event-specific electrical power assumptions (AC/DC), single failures, and the need for operator actions necessary to mitigate non-LOCA design basis events. The staff intends to place a similar L/C on the currently submitted Non-LOCA LTR revision. As such, NuScale is requested to propose markups to remove or sufficiently reword any and all statements that explicitly or implicitly discuss operator actions. For example, the statements could be reworded to clearly identify this assumption within the context of event escalation. For example, if the initiating event is a small leak/break it may be assumed the leak/break does not get worse during the transient timeframe.
Operator actions are considered to mitigate design basis events if the action is relied on in the analysis within the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following the initiating event to improve the results relative to the applicable figures of merit for a particular set of initial conditions.
There are sections throughout the LTR that contain such statements, and the staff has specifically noted the ones below that are necessary to be removed or significantly reworded (if reworded it is necessary to mention that any credit must be sought through specific NRC approval via the design review for such credit). If there are other such statements in other sections, they should also be addressed:
TR Section 4.3.6 states that input to radiological analysis may assume valve isolation times based on MPS response or operator actions, as operator actions may be taken to prevent abnormal operating events from resulting in more severe events.
TR Section 7.1.7 states that operator actions credited for the non-LOCA transient analyses NuScale Nonproprietary NuScale Nonproprietary
are typically justified and consistent with plant operating procedures. For an NPM, there are no occasions where operator action is credited for event mitigation by the non-LOCA transient analyses. Operator actions taken to prevent abnormal operating events from resulting in more severe events are excluded from consideration. For example, very small leaks of reactor coolant from the CVCS that do not result in automatic reactor trip for more than 30 minutes are considered an abnormal operating event where operators are expected to identify and isolate the leak before it results in a more severe event..
TR Section 7.2.16 states that surveillance of the pool boron concentration at appropriate intervals prevents prevent boron dilution in Mode 5. This implies that the operator needs to identify and isolate the source of the diluted water.
Response
TR-0516-49416, Revision 4, Non-Loss-of-Coolant Accident Analysis Methodology, is revised as indicated in the attached markups.
NuScale Nonproprietary NuScale Nonproprietary
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5
© Copyright 2024 by NuScale Power, LLC 48 accident radiological analysis. The conservative bias directions for the transient analysis input to the accident radiological analysis are:
Maximum integrated mass release outside of containment prior to isolation of the RCS mass release.
Basis: For a constant radionuclide concentration in the RCS, the greater the mass released outside of containment, the more severe the radiological consequences.
Maximum integrated mass release between time of reactor trip and time of isolation of the RCS mass release.
Basis: Reference 8 describes how iodine spiking is accounted for in the accident radiological analyses. During a transient progression, changes in reactor power, RCS average temperature, or RCS pressure could result in iodine spiking, changing the radionuclide concentration in the RCS. Consistent with Standard Review Plan Section 15.6.2 (Reference 9), for accident radiological calculations assuming a coincident iodine spike (iodine spike occurring during the event), the iodine spiking is assumed to begin at the time of reactor trip as the result of the reactor shutdown or depressurization of the primary system. In some cases, particularly for smaller breaks in RCS piping, the time between reactor trip and isolation of the break flow could be extended compared to larger break sizes. The increased time between reactor trip and isolation increases the time of mass release when the RCS radionuclide inventory reflects iodine spiking.
Audit Question A-NonLOCA.LTR-43 For each transient analysis case identified, the input provided for accident radiological analysis includes:
Time of reactor trip if it is calculated to occur
Isolation time, at which point release of RCS fluid outside of containment is stopped. The isolation may be due to MPS response or due to operator action (as identified in Section 7.1.7, operator actions may be taken to prevent abnormal operating events from resulting in more severe events).
RCS fluid mass release outside of containment as a function of time
System transient response parameters such as Reactor power as a function of time RCS average temperature as a function of time RCS pressure as a function of time Secondary side feedwater and steam flow rates as a function of time As an alternative to transient analysis, the accident radiological analysis can use bounding values for both mass release and isolation times. An example bounding approach is provided as follows. If the MPS of an NPM design includes reactor trip and isolation setpoints based on pressurizer level, these setpoints effectively limit the RCS inventory that can be released to the difference between a high pressurizer level and the pressurizer level associated with the isolation, and limit the time between
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5
© Copyright 2024 by NuScale Power, LLC 531 reverse flow is experienced during a break in the feedwater system piping.
(Section 7.1.4)
Audit Question A-NonLOCA.LTR-43 Operator actions credited for the non-LOCA transient analyses are typically justified and consistent with plant operating procedures. For an NPM, there are no occasions where operator action is credited for event mitigation by the non-LOCA transient analyses. Operator actions taken to prevent abnormal operating events from resulting in more severe events are excluded from consideration. For example, very small leaks of reactor coolant from the CVCS that do not result in automatic reactor trip for more than 30 minutes are considered an abnormal operating event where operators are expected to identify and isolate the leak before it results in a more severe event.
7.2 Event Specific Methodology The non-LOCA event simulations are performed using conservative methodologies.
Pertinent event-specific methodologies, as well as inputs and results for a representative NPM for non-LOCA event simulations are presented herein, and compared with the regulatory acceptance criteria listed in Table 7-4. Section 4.1 contains additional discussion of Chapter 15 design basis events and acceptance criteria.
All acceptance criteria are considered for each event, and the criteria with the potential for being challenged are identified and evaluated in further detail (i.e., overcooling events do not challenge the acceptance criterion for primary side pressure, but may challenge the CHFR acceptance criteria). An event-specific parameter that is relevant to the acceptance criterion may be described as challenging in the event-specific summary, however, it is recognized that the parameter may not present the worst challenge for any event.
Table 7-4 Regulatory acceptance criteria Description AOO Criteria IE Criteria Accident Criteria RCS pressure 110% of design 120% of design 120% of design SG pressure 110% of design 120% of design 120% of design CHFR(1)
> Limit Note (3)
Note (3)
Maximum fuel centerline temperature(1)
Limit Note (3)
Note (3)
Containment integrity(2)
< Limits
< Limits
< Limits Escalation of an AOO to an accident (AOO) or consequential loss of system functionality (IE or accident)?
No No No Dose(1)
Normal operations
< Limit
< Limit
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5
© Copyright 2024 by NuScale Power, LLC 618 determine the shutdown margin available after isolation of the DWS, and the time at which the shutdown margin would be lost if the dilution source is not terminated.
Mode 4 (Transition)
In this mode of plant operation, all CVCS connections to an NPM are disconnected, isolated, or locked out. Thus, the possibility of a design-basis inadvertent decrease in boron concentration is precluded.
Mode 5 (Refueling)
Audit Question A-NonLOCA.LTR-43 In this mode of plant operation, the Technical Specifications require the pool boron concentration to be sufficient to have appropriate shutdown margin. For some NPM designs, the Technical Specifications also require the pool level to be maintained within a narrow range. Surveillance of the boron concentration, and level if applicable, of the refueling pool is performed at appropriate intervals is expected to prompt operator actions in accordance with Technical Specifications during an inadvertent dilution of the pool; such operator actions are not credited.to prevent significant inadvertent dilution from flow paths to the reactor pool, or proximate water sources such as fire mains or feedwater piping.
The relevant acceptance criteria, SAF, and LOP scenarios are listed in Table 7-72.
7.2.16.2 Acceptance Criteria Evaluation of the most challenging case(s) relative to the acceptance criteria is presented in Table 7-73.
Table 7-72 Acceptance criteria, single active failure, loss of power scenarios - inadvertent decrease in boron concentration Acceptance Criteria / Single Active Failure / Loss of Power Scenarios of Interest Discussion MCHFR MCHFR is challenged during this reactivity anomaly event. (Reactivity insertion rates are insufficient to challenge fuel centerline temperature.)
No single failure The challenging cases typically occur when all equipment is operational.
No loss of power The challenging cases typically occur when AC power is available to the CVCS equipment for the event duration.