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2019/03/20 Nuscale SMR DC RAI - Request for Additional Information No. 520 Erai No. 9642 (16)
	ML19079A363
Person / Time
Site:	NuScale
Issue date:	03/20/2019
From:	; NRC
To:	; NRC/NRO/DLSE/LB1
References
	Download: ML19079A363 (20)
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Paragraph (a)(11) of 10 CFR 52.47 and paragraph (a)(30) of 10 CFR 52.79 state that a design certification (DC) applicant and a combined license (COL) applicant, respectively, are to propose technical specifications (TS) prepared in accordance with 10 CFR 50.36 and 50.36a. 10 CFR 50.36 sets forth requirements for TS to be included as part of the operating license for a nuclear power facility. Generic TS Subsection 3.4.3 Action D states: D. Containment flooding initiated while RCS temperature greater than allowed by PTLR. D.1 Be in MODE 2. l Immediately AND D.2 Be in MODE 3 below the PTLR RCS temperature limit. l 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> AND D.3 Determine RCS is acceptable for continued operation. l Prior to entering MODE 2 from MODE 3 The staff requests additional information about Action D, regarding whether this provision is the appropriate means of addressing prevention and mitigation of the postulated inadvertent actuation of the CFDS to flood the containment vessel with RCS temperature above the RCS temperature limit in the PTLR. Condition D is an unusual Condition, which is based on an inadvertent actuation of the CFDS in MODE 1, 2, or MODE 3 above the PTLR limit on RCS temperature for containment flooding having occurred. Revision 2 of DCA part 2, Tier 2, FSAR Section 9.3.6.2.2 describes the CFDS supply/drain isolation valve interlock with RCS hot temperature: Flooding and draining an individual CNV is conducted through the same CNV penetration. The CFDS pump operation is automatically prevented if the CFDS isolation valve to more than one NPM is open, and valve operation to other NPMs is prevented once the CFDS pump aligned to an NPM and the pump is in service. In addition, for the selected NPM, the CFDS module isolation valve cannot be opened and CFDS pump start is prevented if RCS wide range hot leg temperature is greater than 350°F. These features, coupled with administrative controls in plant procedures, prevent inadvertent CFDS makeup to an operating NPM.

Revision 2 of DCA part 2, Tier 2, FSAR Section 9.3.6.2.2 describes normal operation of the CES and CFDS; in a passage labled "Containment Flooding in Preparation for Refueling," states in part: When the CFDS is used to flood a CNV, one of the CFDS pumps is aligned to take suction from the reactor pool and discharge to the selected NPM. To minimize thermal stress on NPM components, flooding is initiated only after temperatures for the NPM being flooded are below a specified maximum temperature and reactor pool bulk temperature is above a specified minimum temperature. To ensure that component temperature limits are not exceeded and to prevent inadvertent flooding of an operating NPM, the selected CFDS module isolation valve cannot be opened and CFDS pump start is prevented if the selected NuScale Power Module RCS wide range hot leg temperature is greater than 350° F. The CFDS flow path for flooding a CNV includes connections for a temporary skidmounted heater if an off-normal condition requires flooding a CNV with elevated RPV temperatures.

The CFDS alignment for flooding requires that the CFDS containment isolation valves and the CFDS module isolation valve for the NPM being flooded are open. Both CFDS pumps are started with the CNV at atmospheric pressure and CFDS performance is monitored by system flow rate, pressure, and temperature, and CNV level instrumentation. The CNV is flooded approximately to the elevation of the RPV pressurizer baffle plate. Automatic action shuts off the operating CFDS pumps and closes the NuScale Power Module CES isolation valve when the preset water level in the CNV is reached, as determined by CNV level instrumentation. At completion of flooding operation, the CFDS containment isolation valves and the CFDS module isolation valve for the NPM being flooded are closed. Revision 2 of DCA part 2, Tier 2, FSAR Section 9.3.6.3 describes inadvertent actuation of the CFDS to flood containment event: Inadvertent flooding of the CNV for an NPM that is at power or not below the temperature or pressure required for flooding is prevented by system interlocks and administrative controls within plant procedures. Each NPM is isolated from the CFDS by three valves in series, the NPM isolation valve and the two CFDS containment isolation valves. Revision 2 of DCA part 2, Tier 2, FSAR Section 15.1.6.1, Loss of Containment Vacuum/Containment Flooding - Identification of Causes and Accident Description, states in part: The reactor component cooling water system (RCCWS) provides heat removal to the control rod drive system. The RCCWS supplies RCCW to CNTS that then conducts RCCW to CRDS piping that passes through containment to provide this function. If piping containing RCCW were to leak or rupture inside the CNV, a containment flooding event would occur. Other potential containment flooding sources include: feedwater containing line break, main steam containing line break, CVCS fluid containing line break, high point vent fluid containing pipe break, and RCCWS fluid containing line break. The feedwater fluid containing line break event is evaluated in Section 15.2.7, the SLB event is evaluated in Section 15.1.5, and the CVCS fluid containing line break is evaluated in Section 15.6.2. The RCCWS fluid line break is a more limiting containment flooding event than a high point vent fluid pipe because it has a temperature lower than the containment saturation temperature. If the lower temperature RCCWS fluid line ruptures, there would be no immediate boiling, preventing the high containment pressure limit from being reached. The flooding of the CNV could cause an increase in heat transfer from the RPV to containment, cooling the RCS. As the RCS cools, reactor power increases due to the negative moderator coefficient. This unexpected rise in core power would decrease the MCHFR, and lead to an over pressurization of the RPV. 1. Regarding the above quotation from FSAR Section 9.3.6.2.2: 1.1 The first paragraph indicates one CFDS pump is aligned for flooding the CNV; the second paragraph indicates that two CFDS pumps are used, and that the pumps stop when the "specified" CNV water level is reached (based on CNV level instrumentation, which is also used for initiating ECCS - MPS Function 3.3.1.23.a). The applicant is requested to clarify whether one or two pumps are used, and state the design single pump flow rate, and the combined pump flow rate.

1.2 In the first paragraph, what are the approximate RCS hot temperature and reactor pool water temperature limits referred to in: "...temperatures for the NPM being flooded are below a specified maximum temperature and reactor pool bulk temperature is above a specified minimum temperature"?

1.3 In the italicized passage in the first paragraph, explain whether the CFDS pump is prevented from starting because of the RCS hot temperature interlock signal, or because the CFDS NPM isolation valve is closed, and its position indication is interlocked with the pump control circuit.

1.4 The applicant is requested to compare the assumed 40°F RCCW temperature and the assumed 1320 gpm break flow rate from two operating RCCWS pumps to the (? gpm) flow rate from two operating CFDS pumps into containment with a reactor pool source temperature of 110°F, and describe whether the RCCWS pipe break inside containment AOO analysis would bound the analysis of an inadvertent actuation of the CFDS to flood containment event.

1.5 The applicant is requested to explain how and why the CES isolation valve automatically closes when the preset water level in the CNV is reached while flooding the CNV. 2. The applicant is requested to (a) explain why the inadvertent actuation of the CFDS to flood containment event is not considered to be an anticipated operational occurrence (AOO) (See Section 15.1.6 "Loss of Containment Vacuum/Containment Flooding"); and (b) discuss whether operator error in aligning the CFDS to the wrong NPM combined with failure of nonsafety-related interlock on the CFDS NPM isolation valve is less likely than a pipe break inside containment in the RCCWS (AOO?), main steam system (Postulated Accident (PA)), feedwater system (PA), or CVCS (PA), which are categorized as indicated by Table 15.0-1, "Design Basis Events"?

As quoted above, Section 15.1.6.1, page 15.1-25 includes the following: ...The RCCWS fluid line break is a more limiting containment flooding event than a high point vent fluid pipe because it has a temperature lower than the containment saturation temperature. If the lower temperature RCCWS fluid line ruptures, there would be no immediate boiling, preventing the high containment pressure limit from being reached. The flooding of the CNV could cause an increase in heat transfer from the RPV to containment, cooling the RCS. As the RCS cools, reactor power increases due to the negative moderator coefficient. This unexpected rise in core power would decrease the MCHFR, and lead to an over pressurization of the RPV. A loss of containment vacuum event is categorized as an AOO. Typically, pipe system failures are categorized as accidents, but the containment flooding event is conservatively categorized as an AOO.

3. Explain why there should not be an SR to verify the two CFDS containment isolation valves are closed with dc power disconnected or isolated from the solenoid, to prevent engaging the hydraulic system that opens the valves, until RCS hot temperature is less than or equal to 350°F - the PTLR limit? 4. Will the 350°F approximate upper limit for RCS hot temperature allowed by the PTLR to initiate containment flooding be affected by the change in the LCO 3.5.3 upper temperature limit of the reactor pool from 140°F to 110°F? How does this RCS hot temperature upper limit vary with the temperature of the reactor pool water source of the CFDS? 5. Should manual opening of the two CFDS CIVs be blocked unless a permissive signal exists from a new MPS Permissive Function based on wide range RCS hot temperature channels? Or justify why the existing RCS temperature nonsafety-related interlock to prevent opening the nonsafety-related (air or motor?) isolation valve in the CFDS supply line to containment (upstream of CFDS CIVs) using the manual open control function (module control system) provides adequate protection by precluding this event from occurring.

Paragraph (a)(11) of 10 CFR 52.47 and paragraph (a)(30) of 10 CFR 52.79 state that a design certification (DC) applicant and a combined license (COL) applicant, respectively, are to propose technical specifications (TS) prepared in accordance with 10 CFR 50.36 and 50.36a. 10 CFR 50.36 sets forth requirements for TS to be included as part of the operating license for a nuclear power facility. After November 6, 2018, public meeting conference call with NuScale, the staff decided to consider a new RAI question regarding the Bases for Subsection 3.3.1 clearly stating that the Setpoint Program (SP) and the Channel Calibration surveillance requirement in Subsection 3.3.1 provide TS control of interlock and permissive settings. 1. Since the sensors and transmitters for process variables used by the RTS and ESFAS are also used to generate the interlock and permissive signals, a Channel Calibration of an MPS sensor and transmitter satisfies the calibration requirement for the shared interlock sensor and transmitter. However, it is unclear to the staff whether the settings for interlock activation and deactivation are determined using the setpoint methodology specified by the Setpoint Program (SP) and are verified to be set correctly in the SFM as a part of the Channel Calibration of each associated MPS Function. In Revision 2 of DCA part 4, the Applicable Safety Analyses, LCO, and Applicability sections of Subsection B 3.3.1 also state: ...The combination of the continuous self-testing features of the MPS and the CHANNEL CALIBRATION specified by SR 3.3.1.4 verify the OPERABILITY of the interlocks and permissives. The applicant is requested to confirm that the intended meaning of this statement is that interlock settings are controlled by the SP, and are verified during Channel Calibration. 2. The staff notes that Revision 2 of DCA part 4, Subsection B 3.3.1, page B 3.3.1-19, regarding discussion of High Power Range Positive and Negative Rate - Reactor Trip and Demineralized Water System Isolation, states in part:

...The SFM logic unit performs calculations to determine the rate of change and compares the result to a setpoint. The trip provides protection against core damage and protects the reactor coolant pressure boundary (RCPB) during the following events:

Inadvertent decrease in boron concentration in the RCS; and

Control Rod Misoperation. These trips provide protection from the effects of transients that occur at power levels above the N-2H interlock. The High Positive and Negative Power Range Rate trips are automatically bypassed below the N-2H interlock and automatically enabled above the N-2H interlock. Actual setpoints are established in accordance with the Setpoint Program.

2.a. The applicant is also requested to revise this discussion to make clear that the sentence in italics above means that the SP also governs the actual settings of the interlocks and permissives. 2.b. The staff notes that a similar sentence is provided on page B 3.3.1-19 in the discussion of High Power Range Linear Power - Reactor Trip and Demineralized Water System Isolation, but is not provided on page B 3.3.1-20 in the discussion of High Intermediate Range Log Power Rate - Reactor Trip and Demineralized Water System Isolation, nor for any other interlock enabled MPS Instrumentation Function Bases discussion in the Applicable Safety Analyses, LCO, and Applicability sections. The applicant is also requested to revise the Bases so that the relationship of the MPS instrumentation Functions, and their bypassing or enabling interlocks and permissives, to the SP controls and Channel Calibration Surveillances is clear. 3. The applicant is requested to revise SR 3.3.1.4 to explicitly require the Channel Calibration to be performed in accordance with Specification 5.5.10, Setpoint Program, as follows (mark up of Revision 2 of DCA part 4, SR 3.3.1.4): Perform CHANNEL CALIBRATION on each required channel listed in Table 3.3.1-1 in accordance with Setpoint Program. Since SR 3.3.1.1 (Channel Check) and SR 3.3.1.4 (Channel Calibration) apply to every MPS instrument Function listed in Table 3.3.1-1, the applicant is requested to consider whether the phrase "on each required channel listed in Table 3.3.1-1" is needed to understand which MPS instrument Functions require Channel Check, and also Channel Calibration in accordance with the Setpoint Program. 4. The applicant is requested to revise Specification 5.5.10, paragraph b, to include either the revision number or the document date of the NRC approved version of TR-0616-49121-P, "NuScale Instrument Setpoint Methodology."

Paragraph (a)(11) of 10 CFR 52.47 and paragraph (a)(30) of 10 CFR 52.79 state that a design certification (DC) applicant and a combined license (COL) applicant, respectively, are to propose technical specifications (TS) prepared in accordance with 10 CFR 50.36 and 50.36a. 10 CFR 50.36 sets forth requirements for TS to be included as part of the operating license for a nuclear power facility.

It is customary and prudent for a design certification application to provide generic TS Bases along with the generic TS, which are required to be provided in a design certification application by 10 CFR 52.47 and 10 CFR 50.36(a)(2). The Bases should be consistent with the proposed design.

In the LCO 3.0.4 Bases, the applicant is requested to consider the following staff suggested NuScale design-specific paragraph change, as follows (see Rev 2 of DCA Part 4, page B 3.0-7: The provisions of LCO 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS. In addition, the provisions of LCO

3.0.4 shall

not prevent changes in MODES or other specified conditions in the Applicability that result from any unit shutdown. In this context, a unit shutdown is defined as a change in MODE or other specified condition in the Applicability associated with transitioning from MODE 1 to MODE 2, and MODE 2 to MODE 3 and not PASSIVELY COOLED, and not PASSIVELY COOLED to PASSIVELY COOLED. In the SR 3.0.4 Bases, the applicant is requested to consider the following staff suggested NuScale design-specific paragraph change, as follows (see Rev 2 of DCA Part 4, page B 3.0-21: The provisions of SR 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS. In addition, the provisions of LCO

3.0.4 shall

not prevent changes in MODES or other specified conditions in the Applicability that result from any unit shutdown. In this context, a unit shutdown is defined as a change in MODE or other specified condition in the Applicability associated with transitioning from MODE 1 to MODE 2, and MODE 2 to MODE 3 and not PASSIVELY COOLED, and not PASSIVELY COOLED to PASSIVELY COOLED. The staff identified these apparent oversights while verifying the changes in response to RAI 157-9033, Question 16-15 had been incorporated in Subsection B 3.0 of Revision 2 of DCA Part 4.

Paragraph (a)(11) of 10 CFR 52.47 and paragraph (a)(30) of 10 CFR 52.79 state that a design certification (DC) applicant and a combined license (COL) applicant, respectively, are to propose technical specifications (TS) prepared in accordance with 10 CFR 50.36 and 50.36a. 10 CFR 50.36 sets forth requirements for TS to be included as part of the operating license for a nuclear power facility. Parts of this question are a followup of RAI 197-9051, Question 16-28. The staff requests the applicant to address the following concerns not already addressed by the planned supplemental response to RAI 197-9051, Question 16-28: (1) To ensure the SP will govern all Channel Calibration SRs, each Channel Calibration Surveillance statement needs to append the phrase "in accordance with the Setpoint Program."

(2) LCO 3.3.4 needs to specify a Channel Calibration for the Class 1E isolation devices associated with the manual RTS and ESF actuation Functions.

(3) The applicant needs to provide additional justification for why the surveillance column Notes for SR 3.3.1.5, SR 3.3.2.3, and SR 3.3.3.3 are needed. Specifically, address the expected operational restrictions or burdens that would be avoided by invoking the Note. Also, explain how the action requirements would be applied if an associated Class 1E isolation device is known to be unable to open on an OC or UV condition for an MPS Function, an RTS Function, an ESFAS Function, and a manual Function.

(4) The applicant needs to address the expected operational restrictions that would be avoided by invoking the exception to meeting the automatic actuation verification Survellance for each valve and trip breaker specified by the SRs quoted below in the background discussion.

(5) In STS, since an Actions table Note is usually used to specify an allowance to open (or close) a valve (or circuit breaker), which is closed (or open) to comply with a Required Action, provided the valve is operated using administrative controls (which are usually defined and described in the Bases discussion of the Note), the applicant needs to explain the need for specifying such an exception in a Surveillance statement, such as proposed in SR 3.1.9.2, SR 3.4.6.3, SR 3.6.2.2, SR 3.6.2.3, and SR 3.6.2.4; or in a surveillance column Note, such as proposed in SR 3.3.3.2, SR 3.3.3.4, and SR 3.4.6.2. (6) The applicant needs to resolve the apparent error noted below in the background discussion about listing LCO 3.5.2, LCO 3.7.1, and LCO 3.7.2 in the discussion of TSTF-541 in Table C-1 of RCDR Revision 1. Background Discussion:

italics o[ ------------------------------NOTE------------------------------

Not required to be met for dampers and valves locked, sealed or otherwise secured in the actuated position.

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