ML18289A751
| ML18289A751 | |
| Person / Time | |
|---|---|
| Site: | NuScale |
| Issue date: | 10/16/2018 |
| From: | NRC |
| To: | NRC/NRO/DLSE/LB1 |
| References | |
| Download: ML18289A751 (14) | |
Text
NuScaleDCRaisPEm Resource From: Cranston, Gregory Sent: Tuesday, October 16, 2018 1:32 PM To: Request for Additional Information Cc: Lee, Samuel; Cusumano, Victor; Harbuck, Craig; Tesfaye, Getachew; Chowdhury, Prosanta; NuScaleDCRaisPEm Resource
Subject:
Request for Additional Information No. 506 eRAI No. 9614 (16.1.1)
Attachments: Request for Additional Information No. 506 (eRAI No. 9614).pdf Attached please find NRC staffs request for additional information (RAI) concerning review of the NuScale Design Certification Application.
Please submit your technically correct and complete response within 60 days of the date of this RAI to the NRC Document Control Desk.
If you have any questions, please contact me.
Thank you.
1
Hearing Identifier: NuScale_SMR_DC_RAI_Public Email Number: 545 Mail Envelope Properties (BN1PR09MB02588A8182A3DAC8F953200890FE0)
Subject:
Request for Additional Information No. 506 eRAI No. 9614 (16.1.1)
Sent Date: 10/16/2018 1:31:42 PM Received Date: 10/16/2018 1:31:57 PM From: Cranston, Gregory Created By: Gregory.Cranston@nrc.gov Recipients:
"Lee, Samuel" <Samuel.Lee@nrc.gov>
Tracking Status: None "Cusumano, Victor" <Victor.Cusumano@nrc.gov>
Tracking Status: None "Harbuck, Craig" <Craig.Harbuck@nrc.gov>
Tracking Status: None "Tesfaye, Getachew" <Getachew.Tesfaye@nrc.gov>
Tracking Status: None "Chowdhury, Prosanta" <Prosanta.Chowdhury@nrc.gov>
Tracking Status: None "NuScaleDCRaisPEm Resource" <NuScaleDCRaisPEm.Resource@nrc.gov>
Tracking Status: None "Request for Additional Information" <RAI@nuscalepower.com>
Tracking Status: None Post Office: BN1PR09MB0258.namprd09.prod.outlook.com Files Size Date & Time MESSAGE 369 10/16/2018 1:31:57 PM Request for Additional Information No. 506 (eRAI No. 9614).pdf 482556 Options Priority: Standard Return Notification: No Reply Requested: No Sensitivity: Normal Expiration Date:
Recipients Received:
Request for Additional Information No. 506 (eRAI No. 9614)
Issue Date: 10/16/2018 Application
Title:
NuScale Standard Design Certification 048 Operating Company: NuScale Power, LLC Docket No.52-048 Review Section: 16 - Technical Specifications Application Section: 16.1.1; TS Section 1.1; and Subsections 3.3.1, 3.3.3, 3.4.3, 3.6.1, 3.6.2, 3.7.1, 3.7.2 QUESTIONS 16-50 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.
The staff noticed there is no RTS RESPONSE TIME definition in Section 1.1 of the generic TS, even though SR 3.3.2.2 requires verifying RTS RESPONSE TIME is within limits with a 24 month Frequency. DCA Rev 1, generic TS Section 1.1, however, does include an ESF RESPONSE TIME definition, but does not use the term anywhere in the generic TS; especially not in:
SR 3.3.1.3 ("Verify [each MPS instrumentation Function] channel RESPONSE TIME is within limits. l 24 months"),
SR 3.3.2.2 ("Verify [each] RTS [Logic and Actuation Function] RESPONSE TIME is within limits. l 24 months"), and SR 3.3.3.2 ("Verify [each ESF Logic and Actuation Function] required RESPONSE TIME is within limits. l 24 months")
During the discussion of this issue with NuScale in a public conference call on September 4, 2018, NuScale stated it was revising the response time requirements in generic TS Revision 2 (anticipate it being issued in October 2018) in generic TS Section 1.1, and Subsections 3.3.1, 3.3.2, and 3.3.3, by removing the definition of ESF RESPONSE TIME and the defined terms RTS RESPONSE TIME and ESF RESPONSE TIME.
The applicant indicated that the response time testing is described in the NRC approved HIPS topical report, and that the generic TS are being revised to be consistent with it. However, it did not appear that the I&C branch had been made aware of the impending changes. It is also unclear how the generic TS Bases will be revised, except that they will be enhanced to more clearly describe the response time testing.
The applicant is requested to provide justification for not including response time defined terms and their definitions in generic TS Section 1.1, and in response time surveillance requirements in generic TS Section 3.3.
16-51 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.
The proposed PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) definition matches the W-STS definition except that it omits the W-STS definitions phrase, and the low temperature overpressure protection arming temperature, which is also not included in the W-AP1000-STS PTLR definition; this is because the AP1000 design uses the relief valves in the normal residual heat removal system suction line for LTOP and has no valve operator to arm at a particular RCS temperature. It appears that based on this, the applicant concludes that this phrase is not applicable. However, the NuScale LTOP functionality of the three reactor vent valves (RVVs) is automatically enabled by the wide range RCS cold temperature interlock T-1 (2 of 4 channels LTOP enable temperature specified in the PTLR, approximately 325°F). The T-1 interlock LTOP enabling temperature appears analogous to an LTOP arming temperature as used in the W-STS, which is based on a typical LTOP system design, such as the design implemented at Vogtle Electric Generating Station, Units 1 and 2. Therefore, the staff questions omission of an equivalent phrase, such as and the low temperature overpressure protection enable temperature, from the NuScale GTS PTLR definition. The applicant is requested to consider revising the PTLR definition in generic TS Section 1.1 to incorporate reference to the T-1 interlock enabling temperature for the LTOP function of the RVVs for consistency with the W-STS PTLR definition.
16-52 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.
Subsection 3.7.1 Action C states:
C. One flow path with an inner and outer required valve inoperable. l C.1 Isolate the affected main steam line flow path by use of at least one closed and de-activated automatic valve, closed manual valve, or blind flange. l 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> It is not clear to the staff how the GTS define a main steam line flow path. Condition C seems to imply that a flow path is either through both MSIVs, or through both MSIV bypass valves. It is unclear whether Condition C implies a flow path consisting of a combination of an MSIV (inner or outer) and an MSIV bypass valve (outer or inner) respectively.
The staff considers that each of the two main steam lines has four flow paths: an inner MSIV flow path and an associated inner MSIV bypass valve flow path; and an outer MSIV flow path and an associated outer MSIV bypass valve flow path. With this identification of flow paths, the staff suggests phrasing Action C as follows:
C. One flow path main steam line with an inner and outer required automatic isolation valve inoperable. l C.1 Isolate the each affected main steam line flow path by use of at least one closed and de-activated automatic valve, closed manual valve, or blind flange. l 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> The staff notes that the Actions table Note regarding separate condition entry for each MSIV and each MSIV bypass valve (that is, "for each inoperable valve") ought not apply to Condition C, based on the following analysis. In the following discussion, three-digit numbers are used with the following meanings:
First digit => functional designation of valve (1-MSIV; 2-MSIV bypass valve)
Second digit => valve location (1-inner; 2-outer)
Third digit => inoperable function of valve (1-actuation; 2-leakage)
As written in DCA Rev. 1 (and quoted above), Condition C appears to apply when one steam generator's main steam line (outside containment) has:
- 1. Two MSIVs inoperable in the following ways:
1.1.1 an open MSIV inner flow path which is incapable of isolation using the MSIV on either an automatic or manual actuation signal (inner MSIV actuation),
OR 1.1.2 an open or closed inner MSIV with leakage outside the specified limit (inner MSIV leakage).
AND 1.2.1 an open MSIV outer flow path which is incapable of isolation using the MSIV on either an automatic or manual actuation signal (outer MSIV actuation),
OR 1.2.2 an open or closed outer MSIV with leakage outside the specified limit (outer MSIV leakage).
- 2. Two MSIV bypass valves inoperable in the following ways:
2.1.1 an open MSIV bypass valve inner flow path which is incapable of isolation using the MSIV bypass valve on either an automatic or manual actuation signal (inner MSIV bypass valve actuation),
OR 2.1.2 an open or closed inner MSIV bypass valve with leakage outside the specified limit (inner MSIV bypass valve leakage).
AND 2.2.1 an open MSIV bypass valve outer flow path which is incapable of isolation using the MSIV bypass valve on either an automatic or manual actuation signal (outer MSIV bypass valve actuation),
OR 2.2.2 an open or closed outer MSIV bypass valve with leakage outside the specified limit (outer MSIV bypass valve leakage).
Condition C clearly applies to the above combinations of one inner and one outer valve, both valves having the same functional designation, as follows:
1.1.1 + 1.2.1 inner MSIV actuation + outer MSIV actuation 1.1.2 + 1.2.1 inner MSIV leakage + outer MSIV actuation 1.1.1 + 1.2.2 inner MSIV actuation + outer MSIV leakage 1.1.2 + 1.2.2 inner MSIV leakage + outer MSIV leakage 2.1.1 + 2.2.1 inner MSIV bypass actuation + outer MSIV bypass actuation
2.1.2 + 2.2.1 inner MSIV bypass leakage + outer MSIV bypass actuation 2.1.1 + 2.2.2 inner MSIV bypass actuation + outer MSIV bypass leakage 2.1.2 + 2.2.2 inner MSIV bypass leakage + outer MSIV bypass leakage However, it is unclear whether Required Action C.1 (as written in DCA Rev. 1) clearly requires isolating each of the two affected flow paths (inner flow path and outer flow path), within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />; isolating both flow paths is necessary to restore redundant isolation capability for the main steam line. If the intent of Required Action C.1 is to only require isolation of one of the two affected flow paths within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (to restore the isolation function), and rely on the Condition A or B requirement to isolate the other flow path within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (to restore redundancy of the isolation function), the applicant is requested to clarify Required Action C.1, as suggested below.
The following combinations of two inoperable valves having different functional designations in the same main steam line would appear to be addressed by concurrent entry into Condition A for an MSIV and Condition B for an MSIV bypass valve:
1.1.1 + 2.1.1 inner MSIV actuation + inner MSIV bypass valve actuation 1.1.2 + 2.1.1 inner MSIV leakage + inner MSIV bypass valve actuation 1.1.1 + 2.1.2 inner MSIV actuation + inner MSIV bypass valve leakage 1.1.2 + 2.1.2 inner MSIV leakage + inner MSIV bypass valve leakage 1.2.1 + 2.2.1 outer MSIV actuation + outer MSIV bypass valve actuation 1.2.2 + 2.2.1 outer MSIV leakage + outer MSIV bypass valve actuation 1.2.1 + 2.2.2 outer MSIV actuation + outer MSIV bypass valve leakage 1.2.2 + 2.2.2 outer MSIV leakage + outer MSIV bypass valve leakage Since these combinations of valves involve only a loss of isolation function redundancy, the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time to isolate the two affected parallel flow paths (which isolates the main steam line) (or, as implied, to restore each valve to operable status) is appropriate.
It is not clear to the staff whether Condition C (as written in DCA Rev. 1) would apply to the following combinations of two inoperable valves having different functional designations and different locations (one inner valve and one outer valve) in the same main steam line:
1.2.1 + 2.1.1 outer MSIV actuation + inner MSIV bypass valve actuation 1.2.2 + 2.1.1 outer MSIV leakage + inner MSIV bypass valve actuation 1.2.1 + 2.1.2 outer MSIV actuation + inner MSIV bypass valve leakage 1.2.2 + 2.1.2 outer MSIV leakage + inner MSIV bypass valve leakage 1.1.1 + 2.2.1 inner MSIV actuation + outer MSIV bypass valve actuation 1.1.2 + 2.2.1 inner MSIV leakage + outer MSIV bypass valve actuation 1.1.1 + 2.2.2 inner MSIV actuation + outer MSIV bypass valve leakage 1.1.2 + 2.2.2 inner MSIV leakage + outer MSIV bypass valve leakage
By specifying that separate Condition entry is allowed for each MSIV, Condition A could be stated as "One or more MSIV flow paths with the MSIV inoperable." Likewise, Condition B could be stated as "One or more MSIV bypass flow paths with the MSIV bypass valve inoperable."
As stated above, the staff suggests the following phrasing of Action C:
C. One main steam line with an inner and outer automatic isolation valve inoperable. l C.1 Isolate each affected flow path by use of at least one closed and de-activated automatic valve, closed manual valve, or blind flange. l 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> If the intent of DCA Rev. 1 is for Required Action C.1 to only require isolation of one of the two affected flow paths within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, as previously described, then it would be clearer to state Action C as follows:
C. One main steam line with an inner and outer automatic isolation valve inoperable. l C.1 Isolate one affected flow path by use of at least one closed and de-activated automatic valve, closed manual valve, or blind flange. l 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Since Condition C ("One flow path with an inner and outer required valve inoperable." or revised as suggested, "One main steam line with an inner and outer automatic isolation valve inoperable.") applies when two valves in different locations (and having different or the same functional designations) are concurrently inoperable, the staff sees no logical way to apply the Actions table Note, allowing separate Condition entry for each MSIV and each MSIV bypass valve, to Condition C. Therefore, this Note should be moved to the Required Action column of Conditions A and B, and be placed above the designator for Required Action A.1 and the designator for Required Action B.1, and should span the column width (left cell margin to right cell margin; 2.45 inches in width). The Note for Action A should say "Separate Condition entry is allowed for each MSIV flow path." The Note for Action B should say "Separate Condition entry is allowed for each MSIV bypass valve flow path."
The staff also suggests a similar column-spanning Note for the Required Action column of Condition C. The Note would say, "Separate Condition entry is allowed for each main steam line." This suggestion is provided assuming that the intent of DCA Rev. 1 is not to enter LCO 3.0.3 if both main steam lines each have an inner and outer valve inoperable.
The applicant is requested to propose an MSIV Specification consistent with the above suggestions, but which is unambiguous.
16-53 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.
Part A Subsection 3.3.3, ESFAS Logic and Actuation, Action A states:
A. One or more divisions of the LTOP Logic and Actuation Function inoperable. l A.1 Open two reactor vent valves (RVVs). l 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
In the event RCS temperature is above the saturation pressure of the containment vessel, this action would increase containment pressure until the RCS and the containment pressures are in equilibrium. The applicant is requested to (1) point out where such a
transient is discussed in the FSAR (or where it will be added, if it is not described);
(2) clarify in the FSAR discussion whether such a transient is part of the expected normal operation of the unit in Mode 3; and (3) explain why opening the RVVs within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is a prudent action with RCS temperature above the T-2 interlock but below the T-1 interlock.
Part B In DCA Rev. 1, generic TS Subsection 3.4.10 specifies that each RVV that is in the closed position shall be operable, but does not state the implied requirement that all three RVVs shall be closed and operable for LTOP. Three RVVs are required, since two RVVs are necessary to perform the overpressure prevention function; the third RVV accounts for the assumed worst case single active failure of an RVV to open on an LTOP actuation signal. This leads to a rather unconventional construction of the associated Actions. The staff considers a clearer presentation would be for the LCO to explicitly require three RVVs to be closed and OPERABLE for LTOP or at least two RVVs be open. Then the LCO, Applicability, and Actions could be written as shown:
LCO 3.4.10 Three reactor vent valves (RVVs) shall be closed and OPERABLE for LTOP, or two RVVs shall be open each with an OPERABLE vent flow path to the containment vessel.
APPLICABILITY: MODE 3 with wide range RCS cold temperature T-1 interlock, the LTOP enable temperature specified in the PTLR.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One closed RVV A.1 Restore affected RVV to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> inoperable. OPERABLE status.
OR A.2.1 Depressurize the RCS. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> AND A.2.2 Open the affected RVV. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> B. Two closed RVVs B.1 Restore one affected RVV to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> inoperable. OPERABLE status.
OR B.2.1 Depressurize the RCS. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> AND B.2.2 Open two RVVs. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> C. Required Action and C.1 Cease all activity with a Immediately associated Completion potential for increasing RCS time of Condition A or B pressure above the LTOP
CONDITION REQUIRED ACTION COMPLETION TIME not met. valve actuation setpoint.
OR AND Three closed RVVs C.2 Initiate action to depressurize Immediately inoperable. the RCS.
AND C.3 Initiate action to open two Immediately RVVs.
The applicant is requested to consider clarification of the LCO, Applicability, and Action requirements of Subsection 3.4.10, consistent with the above example.
16-54 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.
The staff requests that the licensee rephrase SR 3.7.1.1 as shown, because the FSAR Section 10.3.2.2 descriptions of the secondary MSIV and MSIV bypass valve (MSIBV) provide no details about the design of the valve actuator. Therefore, it is unclear whether this Surveillance applies to the secondary MSIV and MSIBV:
SR 3.7.1.1 Verify required valves the accumulator nitrogen pressure of each safety related MSIV and MSIV bypass valve is pressures are within limits.
The staff requests that the applicant revise Subsection B 3.7.1 to describe the type of valve operator provided for each secondary MS line isolation valve (MSIV and MSIBV), and which SR applies to these valves, since it appears that SR 3.7.1.1 does not apply.
The staff requests that the licensee also rephrase the feedwater isolation valve (FWIV) accumulator pressure surveillance statement with similar edits, as indicated:
SR 3.7.2.1 Verify required the FWIV accumulator nitrogen pressure of each FWIV is pressures are within limits.
These changes are based on the fact that only a subset of the automatic isolation valves required to be operable by LCO 3.7.1 and LCO 3.7.2 have nitrogen accumulator actuators for closing.
16-55 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.
Regarding Surveillances unique to the NuScale design, SR 3.6.2.5 is provided in lieu of conducting an integrated containment leak rate test, which is described and explained in FSAR Section 6.2.6.1; containment penetration leakage rate testing is described in FSAR Section 6.2.6.2; CIV leakage rate testing is described in FSAR Section 6.2.6.3. This Surveillance states:
SR 3.6.2.5 Verify the combined leakage rate for all containment bypass leakage paths is 0.6 La when pressurized to 951 psia.
It is unclear to the staff (1) why this surveillance statement does not identify the pressure criterion of 951 psia as the calculated peak containment internal pressure (Pa); and (2) how this SR relates to SR 3.6.1.1 ("Perform required visual examinations and leakage rate testing in accordance with the Containment Leakage Rate Testing Program."). The applicant is requested to revise SR 3.6.2.5 and its Bases by identifying 951 psia as Pa; and if necessary, by updating this pressure to the most up to date value. The applicant is also requested to explain how SR 3.6.2.5 is related to SR 3.6.1.1, and incorporate this explanation in the SRs section of Subsections B 3.6.1 and B 3.6.2.
16-56 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.
The staff noted that the 30 minute base Frequency (provided in FSAR Table 16.1-1) of SR 3.4.3.1 ("Verify RCS pressure, RCS temperature, and RCS heatup and cooldown rates are within limits specified in the PTLR. l In accordance with the Surveillance Frequency Control Program (SFCP)") is proposed for inclusion in the SFCP. The Note to this Surveillance states "Only required to be performed during RCS heatup and cooldown operations and inservice leak and hydrostatic testing." The staff finds no basis for ever relaxing this 30 minute Frequency during RCS heatup and cooldown operations and inservice leak and hydrostatic testing. The applicant is requested to justify including the Frequency of this Surveillance in the SFCP.
16-57 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.
FSAR Section 16.1.1 describes COL Item 16.1-1 as follows (for completeness, the staff suggests adding a phrase, as shown):
A COL applicant that references the NuScale Power Plant design certification will provide the final plant-specific information identified by [ ] in the generic Technical Specifications and generic Technical Specification Bases.
The applicant is requested to incorporate the suggested phrase in FSAR Section 16.1.1 to clarify that bracketed COL information also resides in the Bases.
16-58 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.
Part 1 The Actions section of the Bases for Subsection 3.7.2 state that "An inoperable FWIV/FWRV may be utilized to isolate the line only if its leak tightness has not been compromised." The Applicability section states "In MODE 3 and not PASSIVELY COOLED, the FWIVs and FWRV[s] are required to be OPERABLE, to support DHRS operability." The ASA section states "The FWIV and FWRV have a specific leakage criteria to maintain DHRS inventory."
To complete its review of SR 3.7.2.3 ("Verify each FWIV and FWRV leakage is within limits. l In accordance with the INSERVICE TESTING PROGRAM"), the staff requests that the applicant explain in the Bases where the FWIV and FWRV leakage limits, including the flowrate value of these limits, are specified.
x Since the design of the FWIV incorporates, in the same valve body, a nozzle check valve that limits DHRS inventory loss, in the event of an upstream feedwater pipe break, by quickly closing (< 1 sec) while the FWIV strokes closed (< 7 seconds), the applicant is requested to describe in the Bases that operability of the check valve is verified as a part of SR 3.7.2.2 ("Verify the closure time of each FWIV and FWRV is within limits on an actual or simulated actuation signal. l In accordance with the INSERVICE TESTING PROGRAM") Testing of this check valve on each feedwater line is described in FSAR, Tier 2, Table 3.9-16, "Valve Inservice Test Requirements per ASME OM Code," Note 9, which states in part:
...The feedwater check valves [(FCVs)] are credited for rapidly acting to the safety function position (closed) to preserve DHRS inventory on a loss of feedwater. The FCVs are normally closed nozzle check valves. The FWIV is credited with providing the primary DHRS/feedwater boundary and has specific leakage criteria. The FCV maintains the DHRS boundary until the FWIV is fully closed and therefore, has no specific leakage criteria. The FCV is located in the same valve body as the FWIV and is located outboard of the two (FWIV located nearest the CNV).
x The design of the FWRV and its associated downstream check valve are not described in detail in Rev. 1 of FSAR, Tier 2, Section 10.4.7.2.2, "Condensate and Feedwater System - System Description - Component Description." FSAR Section 10.4.7.2.2 states in part In off-normal conditions the MPS overrides normal control of the [FWRVs] and can force closure. Each FWRV is designed to fail closed on loss of power or control signal, regardless of the operating mode, and performs a feedwater isolation function as a backup to the FWIV.
x FSAR Section 10.4.7.2.2 also describes the feedwater check valves as follows:
Two check valves are installed in each feedwater line. Both feedwater check valves prevent reverse flow from the steam generators whenever the feedwater system is not in operation and are designed to withstand the forces of closing after a CFWS line rupture. The first check valve is upstream of and integral with the FWIV, providing backflow prevention. The second is downstream of the FWRV and is provided for secondary backflow prevention.
x FSAR Tier 2 Table 10.4-17, "Condensate and Feedwater System Component Design Data," indicates that the FWRV closure type is "air-operated" and its design specification is "in accordance with ASME BP&V Code 2010, 2011 Addenda, Section VIII and Heat Exchanger Institute 2622, 8th Edition."
x FSAR Tier 2 Table 10.4-18, "Condensate and Feedwater System failure Modes and Effects Analysis," states that for the Function of "[controlling] "feedwater flow to the SGs during low flow operations below the feedwater pump VFDs abilities," in the event of a FWRV Failure Mode of: Spurious Opening During startup, and during other low feedwater flow rate operations, the spurious opening of the FWRV results in an increase in flow through the spuriously opened path. If the increase in flow to the SG results in over cooling of the primary side the reactor trips due to high reactor power.
With the inability to control the feedwater flow rate to one of the two steam generators, the DHRS is actuated (on what signal?) and the NuScale Power Module is isolated for decay heat removal.
No safety related portions of the NSSS are affected, as SGs can be isolated by the FWIVs. Decay heat is removed by the DHRS exchanger.
Failure Mode of: Spurious Closing During startup, and during other low feedwater flow rate operations, the spurious closing of the FWRV results in the termination of flow through one of the two SGs. There is no plan to maintain operation if one of the two SGs is unavailable. If the decrease in feedwater flow does not cause a reactor trip due to high pressure on the primary side, the decision is made by operators to trip the reactor due to regulating valve failure.
No safety related portions of the NSSS are affected, as SGs can be isolated by the FWIVs. Decay heat is removed by the DHRS exchanger.
x FSAR Tier 2 Table 10.4-18, "Condensate and Feedwater System failure Modes and Effects Analysis," states that for the Function of "[providing] "
redundant isolation for FWIV actuation events," in the event of a FWRV Failure Mode of: (A) Failure to Close; (B) Slow Closing, or (C) Spurious Opening The FWIVs are the primary method for providing steam generator isolation. There is no effect on reactor safety if the FWRVs fail with the FWIVs operating correctly. Additional protection is provided by the feedwater safety and non-safety check valves.
x FSAR Tier 2 Table 10.4-18, "Condensate and Feedwater System failure Modes and Effects Analysis," states that for the Function of "[providing] "
redundant isolation for safety-related check valve," in the event of a nonsafety check valve (immediately downstream of the FWRV)
Failure Mode of: (A) Failure to Close The safety related check valve is the primary method for maintaining steam generator inventory during a feedwater line break. There is no effect on reactor safety if the [nonsafety] feedwater check valve were to fail with the safety related check valve operating correctly.
Part 2 To complete its review of SR 3.7.1.3 ("Verify each MSIV and MSIV bypass valve leakage is within limits. l In accordance with the INSERVICE TESTING PROGRAM"), the staff requests that the applicant explain in the Bases where the MSIV and bypass valve leakage limits, including the flowrate value of these limits, are specified.
x The design of the Secondary MSIV and Secondary MSIBV and any associated check valves are not described in detail in Rev. 1 of FSAR Section 10.3.2.2, "Main Steam System - System Description -
Component Description."
o FSAR, Tier 2, Table 10.3-1, "MSS Design Data," indicates that the Secondary MSIV is a "12 inch gate valve" with a "hydraulic or pneumatic" actuator system, and a "closure speed" (which is taken to mean valve closure time) of "within 5 seconds"; and that the Secondary MSIBV is a "4 inch [type of valve not stated]" valve with an "air operated" actuator system, and a "closure speed" (which is taken to mean valve closure time) of "within 10 seconds."
16-59 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.
The applicant is requested to explain why DHRS actuation on High Narrow Range Containment Pressure is apparently not required to be operable with RCS Wide Range Hot Temperature < 350°F, which is below the T-3 interlock, even though the NPM is not being PASSIVELY COOLED, because decay heat removal is accomplished using the turbine bypass system, the condensate and feedwater system, and the circulating water system. The Applicability of Function 3.3.1.22b in MODE 3 is "When not PASSIVELY COOLED." However, T-3 automatically bypasses Function 3.3.1.22b below 350°F. See GTS Bases Revision 1, Subsection B 3.3.1, Applicable Safety Analyses, LCO, and Applicability section:
Pages B 3.3.1-14 and -15 regarding Wide Range RCS Hot Temperature Interlock, T-3;
- 1. On decreasing temperature, the T-3 interlock automatically bypasses:
- High Narrow Range Containment Pressure trip for DHR actuation...
Pages B 3.3.1-15 and -16 regarding Containment [Vessel] Level Interlock, L-1:
- 2. On decreasing containment water level or not RT-1 (Reactor Trip Permissive not established [when one or both divisional reactor trip breakers indicate closed]), the L-1 interlock automatically enables the following trip signals for DHR actuation:
- High Narrow Range Containment Pressure Pages B 3.3.1-35 and -36 regarding Narrow Range Containment pressure, in particular, High Narrow Range Containment Pressure Decay Heat Removal System Actuation:
Four High Narrow Range Containment Pressure DHRS channels are required to be OPERABLE in MODES 1 and 2, and MODE 3 without PASSIVE COOLING in operation. In MODE 3 with PASSIVE COOLING in operation, sufficient cooling for decay heat loads is met. In MODES 4 and 5 the reactor is subcritical and passively cooled.
The applicant is requested to capitalized "passively cooled" in the last sentence, because this expression is a defined term.