Text

LLC - Response to NRC Request for Additional Information No. 506 (Erai No. 9614) on the NuScale Design Certification Application
	ML18347A619
Person / Time
Site:	NuScale
Issue date:	12/12/2018
From:	Rad Z; NuScale
To:	; Document Control Desk, Office of New Reactors
References
	RAIO-1218-63828
	Download: ML18347A619 (58)
	v • d • e

RAIO-1218-63828 December 12, 2018 Docket No.52-048 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738

SUBJECT:

NuScale Power, LLC Response to NRC Request for Additional Information No.

506 (eRAI No. 9614) on the NuScale Design Certification Application

REFERENCE:

U.S. Nuclear Regulatory Commission, "Request for Additional Information No.

506 (eRAI No. 9614)," dated October 16, 2018 The purpose of this letter is to provide the NuScale Power, LLC (NuScale) response to the referenced NRC Request for Additional Information (RAI).

The Enclosure to this letter contains NuScale's response to the following RAI Questions from NRC eRAI No. 9614:

16-50

16-51

16-52

16-53

16-54

16-55

16-56

16-57

16-58 The response to RAI question 16-59 will be provided by March 01, 2019.

This letter and the enclosed response make no new regulatory commitments and no revisions to any existing regulatory commitments.

If you have any questions on this response, please contact Carrie Fosaaen at 541-452-7126 or at cfosaaen@nuscalepower.com.

Sincerely, Zackary W. Rad Director, Regulatory Affairs NuScale Power, LLC NuScale Power, LLC 1100 NE Circle Blvd. , Suite 200 Corvalis, Oregon 97330 , Office: 541.360.0500 , Fax: 541.207.3928 www.nuscalepower.com

RAIO-1218-63828 Distribution: Gregory Cranston, NRC, OWFN-8G9A Samuel Lee, NRC, OWFN-8G9A Getachew Tesfaye, NRC, OWFN-8H12 : NuScale Response to NRC Request for Additional Information eRAI No. 9614 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com

RAIO-1218-63828 :

NuScale Response to NRC Request for Additional Information eRAI No. 9614 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com

Response to Request for Additional Information Docket No.52-048 eRAI No.: 9614 Date of RAI Issue: 10/16/2018 NRC Question No.: 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, NuScale Nonproprietary

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.

NuScale Response:

NuScale Response The NuScale design uses a digital module protection system (MPS) to detect and mitigate transients and accidents as described in FSAR chapter 7, Instrumentation and Controls. The MPS design is based on the Highly Integrated Protection System (HIPS) described in TR-1015-18653, Topical Report Design of the Highly Integrated Protection System Platform, Revision 2-A. The report describes the digital protection system features that include independence, redundancy, predictability, repeatability, diversity, and defense in depth.

The HIPS platform takes advantage of the relatively simple architecture and deterministic functionality of older analog protection system designs and implements these concepts using complex digital logic resulting in a protection system that is designed to produce the same outputs for a given set of input signals within well-defined response time limits to allow timely completion of credited actions. These differences led to a complete restructuring of the instrumentation section of the proposed generic technical specifications (TS), including redefinition of terminology and concepts used there. Response time testing is one area that required significant changes.

The historic response time definitions described in the NUREG series of standard technical specifications (STS) were developed and based on analog protection system concepts. They were determined to be inappropriate for use with the NuScale design, and their limited usage with a high potential for misinterpretation was identified if the same or similar terminology was adopted in the NuScale TS.

NuScale Nonproprietary

The details of the design resulted in a different paradigm for verifying that protection features would detect and respond to conditions that require mitigation in a timely manner. The new paradigm divides response time verification into three conceptual portions - sensor inputs, digital protection system (MPS), and logic and actuated components.

This approach was developed because when a signal enters the digital protection system it loses identity as it is processed, combined, and appropriate actuations are initiated. Additionally, the digital signal processing occurs in multiple devices at extremely high speeds compared to classical relay designs. The HIPS topical report describes the processing in detail and defines a means of calculating the limiting upper bound of potential delay in the system. The calculated bounding response times are defined by the details of the system components and interconnections and is a fixed parameter of the design that does not change. The response time will be verified during factory acceptance testing of the MPS as described in associated inspections, tests, analyses, and acceptance criteria listed in Tier 1, Table 2.5-7 of the FSAR.

The self-testing features of the design will notify operators of failures that could impact system function, however degradation of the system response time cannot occur. An OPERABLE MPS has a defined digital response time that does not change and does not require further verification.

With the digital portions of the maximum response time delays in actuation of credited safety features defined, the two remaining portions of the response time of credited systems remain to be addressed by the surveillance requirements.

TS 3.3.1, Module Protection System (MPS) Instrumentation surveillance requirement (SR) 3.3.1.3 requires that the response time for each channel be verified within limits. This is consistent with the common STS approach of requiring the response times to be tested.

However with the integrated MPS instrumentation, the historic definitions become inappropriate because many of the channel response times that are being verified are used to provide both reactor trip system actuation and Engineered Safety Features Actuation System (ESFAS) actuation.

Recognizing that the response times of logic and actuated components downstream of the MPS must be verified, surveillance requirements are specified for ensuring they remain within limits.

As described in the FSAR, actuation of safety components is accomplished by interrupting the electrical power to the component. The measurement verification of response time of output logic and actuation begins at the output of the MPS and continues through the actuated component reaching its safety position. With the exception of electrical breakers that must open to initiate a reactor trip or to de-energize the pressurizer heaters, each safety actuation results in repositioning valves either open or closed.

NuScale Nonproprietary

Actuated valves with response time requirements listed in the TS have their response time measured in accordance with the surveillance requirements listed below. The scope of the testing is clarified in the associated Bases to ensure that testing is measured from the initiation of the actuation signal at the output of the MPS until the valve has achieved its safety position.

This ensures that the entire logic from the MPS through the repositioning of the component is verified to be within limits as shown below.

Actuated Valve Response Time Surveillance Requirements

SR 3.4.6.2 CVCS isolation valves

SR 3.4.10.2 Reactor vent valves (RVV) (for LTOP function)

SR 3.5.1.2 RVV and reactor recirculation valves (for ECCS function)

SR 3.5.2.4 DHRS actuation valves

SR 3.6.2.3 Containment isolation valves

SR 3.7.1.2 Main steam line isolation valves

SR 3.7.2.2 Feedwater isolation valves The MPS delay time illustrated in the figures below is the conservatively limiting delay time calculated using the methodology described in section 7.7 of "Design of the Highly Integrated Protection System Platform," TR-1015-18653-P-A, Rev. 2 (ML17150A509).

NuScale Nonproprietary

Pressurizer heater breaker response time verification is measured in a similar manner. The response time is required by SR 3.3.3.2 because the OPERABILITY of the breakers is fully addressed by the requirements of LCO 3.3.3, ESFAS Logic and Actuation.

Reactor trip breakers are verified to respond within limits using a slightly different paradigm.

Verification of the response time from output of the MPS through opening of the reactor trip breakers is required by SR 3.3.2.2. However the response time through actuation of the components includes the control rod assembly (CRA) insertion time verification that is required by SR 3.1.4.3 as illustrated below.

NuScale Nonproprietary

This response time testing paradigm has been incorporated into Revision 2 of the generic technical specifications and bases for the NuScale design (ML18310A348 and ML18310A349).

While different from historical plant testing schemes it will insure that the credited features required to actuate will do so within the time assumed in the plant analyses. As noted above, the STS definitions were determined inconsistent with the paradigm and therefore are not utilized in the proposed TS.

Impact on DCA:

There are no impacts to the DCA as a result of this response.

NuScale Nonproprietary

Response to Request for Additional Information Docket No.52-048 eRAI No.: 9614 Date of RAI Issue: 10/16/2018 NRC Question No.: 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 WAP1000-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.

NuScale Nonproprietary

NuScale Response:

The NuScale Low Temperature Overpressure Protection 'arming temperature' is established and maintained as specified in TS 5.5.10, Setpoint Program, and described in the Bases for LCO 3.3.1, Module Protection System.

The design of a NuScale reactor is quite different from existing PWRs. The design implementation of LTOP is quite different from other PWRs. Modifying the proposed NuScale generic technical specifications, PTLR contents, design expectations, and operating paradigm to assure consistency with another vendor's design and implementation would be inappropriate.

Impact on DCA:

There are no impacts to the DCA as a result of this response.

NuScale Nonproprietary

Response to Request for Additional Information Docket No.52-048 eRAI No.: 9614 Date of RAI Issue: 10/16/2018 NRC Question No.: 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 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 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 0.00222 hours 1.322751e-5 weeks 3.044e-6 months NuScale Nonproprietary

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 bypassvalve)

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 NuScale Nonproprietary

MSIV leakage).

OR

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 NuScale Nonproprietary

1.1.2 + 1.2.1 inner MSIV actuation + outer MSIV leakage 1.1.2 + 1.2.1 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 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ; 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 0.00222 hours 1.322751e-5 weeks 3.044e-6 months (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 0.02 hours 1.190476e-4 weeks 2.7396e-5 months (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 NuScale Nonproprietary

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 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 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 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 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 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , as previously described, then it would be clearer to state NuScale Nonproprietary

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 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 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.

NuScale Response:

LCO 3.7.1 has been simplified to more clearly reflect the NuScale design and assure that the LCO requirements are consistent with the FSAR description, the credited function, and TSTF-GG-05-01, Revision 1, Writer's Guide for Plant-Specific Improved Technical Specifications.

Each NuScale steam generator (SG) supplies steam to a steam line that passes through one or two safety related valves. These flow paths are above the reactor module as described in FSAR section 6.2. The steam then flows to the steam lines attached to the reactor building where it passes through one or two non-safety related valves and flow paths. This is schematically NuScale Nonproprietary

illustrated below (see FSAR section 6.2, figure 6.2-4 and the main steam system description provided in FSAR section 10.3.2 for additional details.)

§ SR - Safety-Related,

§ MSIV - Main Steam Isolation Valve,

§ MSIBV - Main Steam Isolation Bypass Valve LCO 3.7.1 requires all four valves on each steam line to be OPERABLE. Condition A requires the flow path through any valve that is inoperable to be isolated within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and verified to remain isolated once every 7 days. Condition B applies if a steam line cannot be isolated due to inoperable valves. This condition applies regardless of the specific valves that are inoperable or flow path combinations between a SG and the secondary plant.

The former Note 1 to the Actions table has been relocated to the Condition column so that is only applicable to Condition A. Corresponding changes to the Bases were also made.

Impact on DCA:

The Technical Specifications have been been revised as described in the response above and as shown in the markup provided in this response.

NuScale Nonproprietary

MSIVs 3.7.1 3.7 PLANT SYSTEMS 3.7.1 Main Steam Isolation Valves (MSIVs)

LCO 3.7.1 Two MSIVs and two MSIV bypass valves per steam line shall be OPERABLE.

APPLICABILITY: MODES 1 and 2, MODES 3 and not PASSIVELY COOLED.

ACTIONS

NOTES-----------------------------------------------------------

1. Separate Condition entry is allowed for each inoperable valve.

2. Main steam line flow path(s) may be unisolated intermittently under administrative controls.

CONDITION REQUIRED ACTION COMPLETION TIME A. -----------NOTE------------ A.1 Isolate the affected main 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Separate Condition steam line flow path by use entry is allowed for of at least one closed and each inoperable valve. de-activated automatic

valve, closed manual valve, or blind flange.

One or more required MSIV valves inoperable. AND A.2 --------------NOTES--------------

1. Isolation in high radiation areas may be verified by use of administrative means.

2. Isolation devices that are locked, sealed, or otherwise secured may be verified by use of administrative means.

Verify the affected main Once per 7 days steam line flow path is isolated.

NuScale 3.7.1-1 Draft Revision 3.0

MSIVs 3.7.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. One or more required B.1 Isolate the affected main 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months MSIV bypass valves steam line bypass flow path inoperable. by use of at least one closed and de activated automatic valve, closed manual valve, or blind flange.

AND B.2 NOTES

1. Isolation in high radiation areas may be verified by use of administrative means.

2. Isolation devices that are locked, sealed, or otherwise secured may be verified by use of administrative means.

Verify the affected main Once per 7 days steam line bypass flow path is isolated.

CB. Steam line that cannot CB.1 Isolate the affected main 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months be isolated. One flow steam line. flow path by use path with an inner and of at least one closed and outer required valve de activated automatic inoperable. valve, closed manual valve, or blind flange.

DC. Required Action and DC.1 Be in MODE 2. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND DC.2 Be in MODE 3 and 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months PASSIVELY COOLED.

NuScale 3.7.1-2 Draft Revision 3.0

MSIVs B 3.7.1 B 3.7 PLANT SYSTEMS B 3.7.1 Main Steam Isolation Valves (MSIVs)

BASES BACKGROUND Each main steam line has one safety related MSIV and one non safety MSIV to isolate steam flow when required to support decay heat removal system (DHRS) operation or containment system (CNTS) operation. The safety related MSIV is located outside of, and close to the containment.

Each main steam line also includes a nonsafety related secondary MSIV located downstream of the removable pipe spool between the module and the balance of the steam system. Each of the four MSIVs is provided with a bypass line that contains an associated bypass isolation valve, one safety related and one non safety related at the corresponding MSIVs. A description of the safety related MSIVs is found in FSAR Section 6.2 (Ref. 1). A description of the nonsafety related secondary MSIVs is found in FSAR Section 10.3. (Ref. 2).

The safety related MSIVs and nonsafety related secondary MSIVs, as well as the four normally closed MSIV Bypass Valves, are closed on a steam line isolation signal, Decay Heat Removal System (DHRS) and Containment Isolation System actuations as described in Specification 3.3.1. Each MSIV and MSIV Bypass Valve closes on loss of power.

Closing the associated Steam Generator (SG) MSIVs and MSIV Bypass Valves isolates the Turbine Bypass System and other steam flows from the SG to the balance of plant.

The MSIVs isolate steam flow from the secondary side of the associated SG following a high energy line break and preserves the reactor coolant system (RCS) inventory in the event of a steam generator tube failure (SGTF). The MSIVs and MSIV Bypass Valves also form part of the boundary of the safety related, closed loop, DHRS described in FSAR Section 5.4 (Ref. 3) and applicable requirements are in Specification 3.5.2.

Each steam generator (SG) supplies one main steam line. Each main steam line includes four isolation valves that isolate steam flow to support decay heat removal system (DHRS) operation or containment system function. Two safety-related valves are located outside of and close to the containment. A description of the safety-related MSIVs is found in FSAR Section 6.2 (Ref. 1). Two non-safety related backup isolation valves are located downstream of the removable pipe spool between the module and balance of the main steam system. A description of the nonsafety-related backup MSIVs is found in FSAR Section 10.3. (Ref. 2).

NuScale B 3.7.1-1 Draft Revision 3.0

MSIVs B 3.7.1 BASES BACKGROUND (continued)

The four valves are arranged so that each MSIV is provided with a bypass line that includes a MSIV bypass isolation valve, one safety related and one non-safety related, arranged in parallel with the corresponding MSIVs.

The safety-related MSIVs and non-safety related backup MSIVs, as well as the normally-closed MSIV bypass valves, will receive and close upon receipt of a steam line isolation signal initiated with Decay Heat Removal System (DHRS) and Containment Isolation System actuations as described in Specification 3.3.1. Each of the MSIV and MSIV Bypass Valves is designed to close upon loss of power.

Closing the MSIVs and MSIV bypass valves isolates the Turbine Bypass System and other steam flows from the SG to the balance of plant. The MSIVs isolate steam flow from the secondary side of the associated SG following a high-energy line break and preserves the reactor coolant system (RCS) inventory in the event of a steam generator tube failure (SGTF). The MSIVs and MSIV bypass valves also form part of the boundary of the safety-related, closed-loop, DHRS described in FSAR Section 5.4 (Ref. 3).

APPLICABLE The MSIVs and MSIV Bypass Isolation Valves close to isolate the SAFETY SGs from the power conversion system. Isolation limits ANALYSES postulated releases of radioactive material from the SGs in the event of a SG tube failure (Ref. 4) and terminates flow from SGs for postulated steam line breaks outside containment (Ref. 5). This minimizes radiological contamination of the secondary plant systems and components, and minimizes associated potential for activity releases to the environment, and preserves RCS inventory in the event of a SGTF.

The isolation of steam lines is also required for the operation of the DHRS. Isolation valve closure precludes blowdown of more than one SG, preserving the heat transfer capability of an unaffected SG if a concurrent single failure occurs. The DHRS provides cooling for non-loss-of-coolant accident (non-LOCA) design basis events when normal secondary-side cooling is unavailable or otherwise not utilized. The DHRS removes post-reactor trip residual and core decay heat and allows transition of the reactor to safe shutdown conditions.

The safety-related and nonsafety-related MSIV and MSIV bypass valves satisify Criterion 3 of 10 CFR 50.36(c)(2)(ii).

NuScale B 3.7.1-2 Draft Revision 3.0

MSIVs B 3.7.1 BASES LCO This LCO requires the safety related and non safety related MSIVs and MSIV Bypass Valves in each of the two steam lines to be OPERABLE.

The valves are considered OPERABLE when their isolation times are within limits and they close on an isolation actuation signal and their valve leakage is within limits.

This LCO provides assurance that the safety related and non safety related MSIVs and MSIV Bypass Valves will perform their design safety function to limit consequences of accidents that could result in offsite exposures comparable to the 10 CFR 50.34 limits or the NRC staff approved licensing basis.This LCO requires four isolation valves on each SG steam line to be OPERABLE. This includes safety related and non-safety related MSIVs and MSIV bypass valves in each steam line. The valves are considered OPERABLE when they will close on an isolation actuation signal, their isolation times are within limits, and valve leakage is within limits.

This LCO provides assurance that the safety related and non-safety related MSIVs and MSIV bypass valves will be available to perform their design safety function to limit consequences of accidents that could result in offsite exposures comparable to the 10 CFR 50.34 limits or the NRC staff approved licensing basis.

APPLICABILITY The safety related and non safety related MSIVs and MSIV Bypass Valves must be OPERABLE in MODE 1, 2, and MODE 3 when not PASSIVELY COOLED. Under these conditions, the isolation of the MSIVs ensures the DHRS can perform its design function and the valves provide a barrier to limit the release of radioactive material to the environment.

Closure of the MSIVs also preserves the RCS inventory in the event of a SGTF. Therefore, these valves must be OPERABLE or closed. When these valves are closed they are performing their required function. In MODES 4 and 5, the unit is shutdown, the SGs do not contain significant energy or inventory, and therefore the MSIVs do not perform any credited safety function. The safety related and non-safety related MSIVs and MSIV Bypass Valves must be OPERABLE in MODE 1, 2, and MODE 3 when not PASSIVELY COOLED. Under these conditions, the isolation of the MSIVs ensures the DHRS can perform its design function and the valves provide a barrier to limit the release of radioactive material to the environment. Closure of the MSIVs also preserves the RCS inventory in the event of a SGTF. Therefore, these valves must be OPERABLE or the flow path through the valve isolated. When these valves are closed or their flow path is isolated, the required function has been satisfied. 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.

NuScale B 3.7.1-3 Draft Revision 3.0

MSIVs B 3.7.1 BASES ACTIONS The ACTIONS are modified by two Notes. The first stating that a separate Condition entry is allowed for each inoperable valve. This is acceptable because the Required Actions provide appropriate compensatory actions for each inoperable isolation valve on each steam line. The second note indicating that MSIV 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 With a required MSIV valve inoperable, isolation of the main steam flow path using the MSIV and MSIV Bypass Valves and supported safety functions can no longer accommodate a single failure. The redundant isolation valves in the affected flow path preserve the ability to isolate the steam flow path.

Action A.1 requires isolation of the main steam line within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Some repairs of the valves may be accomplished within the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months period to restore OPERABILITY and exit the LCO. The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months completion time is reasonable because the inoperable isolation valve only affects the capability of one of the two redundant DHR trains to function. Only if a single failure occurs that affects the remaining capability to isolate the steam flow path will the DHR train be affected.

The 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. Alternatively if the main steam line can be isolated by closing the inoperable valve within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months then its function is being performed. The capability to isolate steam flow if a single failure occurs remains unaffected. If the MSIV is inoperable and cannot be closed, then the steam line should be isolated by the other MSIV and associated bypass valve closed and deactivated, closed manual valve, or blind flange. An inoperable MSIV may be utilized to isolate the line only if its leak tightness has not been compromised. The 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is reasonable to adjust unit conditions and take action to isolate the line.

Required Action A.2 is modified by two notes. Note 1 applies to isolation devices located in high radiation areas and allows these devices to be verified closed by use of administrative means. Allowing verification by administrative means is acceptable, since access to these areas is typically restricted. Note 2 applies to isolation devices that are locked, sealed, or otherwise secured in position and allows these devices to be NuScale B 3.7.1-4 Draft Revision 3.0

MSIVs B 3.7.1 BASES ACTIONS (continued) 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.

For inoperable components that are not restored to OPERABLE status prior to the required completion time in Required Action A.1 and now have their flow path isolated, Required Action A.2 is applicable. Action A.2 requires that the flow path be verified isolated on a periodic basis.

The 7 day Completion Time is reasonable based on engineering judgement, valve and system status indications available in the control room, and other administrative controls, to ensure these flow paths are isolated.

This condition applies if one MSIV is inoperable in either or both main steam lines. In this case, capability to automatically isolate the steam flow path is preserved by the inner (closest to Containment) or outer (furthest from Containment) redundant valves. If isolation capability is not maintained because of the combination of multiple isolation valves in the same flow path being inoperable, then Condition C is applicable.

B.1 and B.2 With a required MSIV Bypass line valve inoperable, isolation of the main steam flow path using the MSIV and MSIV Bypass Valves and supported safety functions can no longer accommodate a single failure. The redundant isolation valves in the affected flow path preserve the ability to isolate the steam flow path.

Action B.1 requires isolation of the main steam bypass line within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Some repairs of the valves may be accomplished within the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months period. The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months completion time is reasonable because the inoperable isolation valve only affect the capability of one of the two redundant DHR trains to function. Only if a single failure occurs that affects the remaining capability to isolate the steam flow path will the DHR train be affected.

The 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. Alternatively if NuScale B 3.7.1-5 Draft Revision 3.0

MSIVs B 3.7.1 BASES ACTIONS (continued) the bypass line can be isolated by closing the inoperable valve within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months then its function is being performed. The capability to isolate steam flow if a single failure occurs, remains unaffected. If the MSIV bypass valve is inoperable and cannot be closed, then the steam line should be isolated by the other MSIV and associated bypass valve closed and deactivated, closed manual valve, or blind flange. An inoperable MSIV bypass valve may be utilized to isolate the line only if its leak tightness has not been compromised. The 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is reasonable to adjust unit conditions and take action to isolate the line.

Required Action B.2 is modified by two notes. Note 1 applies to isolation devices located in high radiation areas and allows these devices to be verified closed by use of administrative means. Allowing verification by administrative means is acceptable, since access to these areas is typically restricted. Note 2 applies to isolation devices that are locked, sealed, or otherwise secured in position and allows these devices 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.

For inoperable components that are not restored to OPERABLE status prior to the required completion time in Required Action B.1 and now have their flow path isolated, Required Action B.2 is applicable. Action B.2 requires that the flow path be verified isolated on a periodic basis.

The 7 day Completion Time is reasonable based on engineering judgement, valve and system status indications available in the control room, and other administrative controls, to ensure these flow paths are isolated.

This condition applies if one MSIV bypass is inoperable in either or both main steam lines. In this case, capability to automatically isolate the steam flow path is preserved by the inner (closest to Containment) or outer (furthest from Containment) redundant valves. If isolation capability is not maintained because of the combination of multiple isolation valves in the same flow path being inoperable, then Condition C is applicable.

NuScale B 3.7.1-6 Draft Revision 3.0

MSIVs B 3.7.1 BASES ACTIONS (continued)

C.1 With a flow path with both an inner and outer required valve inoperable, isolation of the main steam flow path using the safety related or non safety related MSIV and MSIV Bypass Valves and supported safety functions can no longer accommodate a single failure. This action applies to both MSIVs and MSIV Bypass valves in the same flow path.

Action C.1 requires isolation of the main steam line flow path by use of at least one closed and deactivated automatic valve, closed manual valve, or blind flange within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . Some repairs of the valves may be accomplished within the 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period. The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months completion time is reasonable because the inoperable isolation valve only affect the capability of one of the two redundant DHR trains to function. Separate entries are allowed if more than one main steam line is affected.

The time is reasonable considering the availability of other means of mitigating design basis events, including Emergency Core Colling System and the low probability of an accident occurring during this time period that would require closure of the specific flow path.

This condition applies if a MSIV and/or MSIV bypass is inoperable in the inner (closest to Containment) set of valves and a MSIV and/or MSIV bypass is inoperable in the outer (furthest from Containment) set of redundant valves. In this case, capability to automatically isolate the steam flow path is compromised.

If the main steam line can be isolated by closing the inoperable valves within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months then its function is being performed. If the MSIV or Bypass valve is inoperable and cannot be closed, then the steam line should be isolated by the other MSIV and associated bypass valve closed and deactivated, closed manual valve, or blind flange. An inoperable MSIV or bypass valve may be utilized to isolate the line only if its leak tightness has not been compromised.

NuScale B 3.7.1-7 Draft Revision 3.0

MSIVs B 3.7.1 BASES ACTIONS (continued)

D.1 and D.2 With Required Actions and associated Completion Times not met, isolation capability of the main steam line(s) is not maintained. The associated DHRS and the ability to isolate postulated releases from the SGs are affected. The unit must be placed in a condition in which the LCO does not apply using Required Action D.1 and D.2.

Required Action D.1 requires the unit must be placed in MODE 2 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Required Action D.2 requires the unit to be in MODE 3 and PASSIVELY COOLED within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

The Completion Times are reasonable based operating activities required to reach these conditions in an orderly manner. The time permits use of normal means to exit the conditions of Applicability. It is also consistent with the Completion Times for an inoperable train of the DHRS.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 This Conditions is modified by a Note stating that a separate Condition entry is allowed for each inoperable 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 assured. 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.

NuScale B 3.7.1-8 Draft Revision 3.0

MSIVs B 3.7.1 BASES ACTIONS (continued)

Action A.1 requires isolation of the inoperable valve flow path within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Some repairs may be accomplished within the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months period to restore OPERABILITY and exit the LCO. The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 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 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.

Alternatively, if the valve flow path can be isolated by closing the inoperable valve within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months then its function is being performed. The capability to isolate steam flow if a single failure occurs remains unaffected.

An inoperable MSIV may be utilized to isolate the flow path only if its leak tightness has not been compromised. The 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is reasonable to adjust unit conditions and take action to isolate the flowpath.

Required Action A.2 is modified by two notes. Note 1 applies to isolation devices located in high radiation areas and allows these devices to be verified closed by use of administrative means. Allowing verification by administrative means is acceptable, since access to these areas is typically restricted. Note 2 applies to isolation devices that are locked, sealed, or otherwise secured in position and allows these devices 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.

For inoperable components that are not restored to OPERABLE status prior to the required completion time in Required Action A.1 and now have their flow path isolated, Required Action A.2 is applicable. Action A.2 requires that the flow path be verified isolated on a periodic basis.

The 7 day Completion Time is reasonable based on engineering judgement, valve and system status indications available in the control room, and other administrative controls, to ensure these flow paths remain isolated.

NuScale B 3.7.1-9 Draft Revision 3.0

MSIVs B 3.7.1 BASES ACTIONS (continued)

B.1 With a steam line that cannot be manually or automatically isolated the supported safety functions can no longer be met. This condition applies when two or more inoperable isolation valves prevent automatic or manual isolation of steam flow from the steam generator. This condition exists when a flow path through the safety related MSIV and MSIV bypass valve exists, and a flow path through the non-safety related secondary MSIV and MSIV bypass valve exists, that cannot be manually or automatically isolated.

For example, one MISV bypass valve inoperable and open, and one non-safety related secondary MSIV inoperable and open could prevent isolation of the steam flow from the associated steam generator. In this condition a steam line flow could exist through the MSIV bypass valve and the secondary MSIV that could not be isolated.

Action B.1 requires isolation of the main steam line by closure of valves so that the safety function of the steam line isolation is accomplished.

Some repairs may be accomplished within the 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period. The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months completion time is reasonable because the inoperable isolation valve only affect the capability of one of the two redundant DHRS trains to function.

The time is reasonable considering the availability of other means of mitigating design basis events, including Emergency Core Colling System and the low probability of an accident occurring during this time period that would require isolation of the steam line.

If the main steam line can be isolated within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months then its safety function is being performed. An inoperable MSIV or bypass valve may be utilized to isolate the steam line only if its leak tightness has not been compromised.

NuScale B 3.7.1-10 Draft Revision 3.0

MSIVs B 3.7.1 BASES ACTIONS (continued)

C.1 and C.2 With Required Actions and associated Completion Times not met, isolation capability of the main steam line(s) is not maintained. The associated DHRS and the ability to isolate postulated releases from the SGs are affected. The unit must be placed in a condition in which the LCO does not apply using Required Action C.1 and C.2.

Required Action C.1 requires the unit must be placed in MODE 2 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Required Action C.2 requires the unit to be in MODE 3 and PASSIVELY COOLED within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

The Completion Times are reasonable based operating activities required to reach these conditions in an orderly manner. The time permits use of normal means to exit the conditions of Applicability. It is also consistent with the Completion Times for an inoperable train of the DHRS.

NuScale B 3.7.1-11 Draft Revision 3.0

Response to Request for Additional Information Docket No.52-048 eRAI No.: 9614 Date of RAI Issue: 10/16/2018 NRC Question No.: 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 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

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 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 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 NuScale Nonproprietary

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:

[see eRAI 9614 question 16-53 (ML18289A51) for proposed LCO]

The applicant is requested to consider clarification of the LCO, Applicability, and Action requirements of Subsection 3.4.10, consistent with the above example.

NuScale Response:

In addition to the RAI as received, the NRC staff provided the following additional discussion during a public meeting on November 6, 2018 (numbering reflects those comments):

32. GTS 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 in Figure I-1. Pending resolution of the staffs concern about the clarity of Subsection 3.4.10, this editorial issue will be tracked as an open item.

NuScale Responses Part A

1) The LTOP function is described in FSAR section 5.2, and FSAR section 3.9 describes the Service Level B cold overpressure protection transient. LTOP actuation results in opening of a vent path between the reactor coolant system (RCS) inside of the reactor vessel and the containment volume. This is similar to, but at reduced temperatures and pressures compared to the ECCS actuation transients from operating conditions described in Chapter 15.

NuScale Nonproprietary

2) An LTOP actuation event is not considered a normal operation - it is a protective system actuation. However the NuScale design includes the availability of PASSIVE COOLING through use of open reactor vent valves (RVV) and reactor recirculation valves (RRV) when conditions permit. This evolution may be manually initiated or otherwise implemented in accordance with operating procedures, and includes the opening of at least two reactor vent valves. Normal operations include de-energization of the reactor module resulting in opening of the RVVs and RRVs when transitioning to a PASSIVELY COOLED configuration before entering MODE 4.

3) The purpose of the LTOP logic and actuation function is to detect and mitigate a low temperature overpressure condition. Opening RVVs above the T-2 interlock but below the T-1 interlock establishes a vent path from the reactor vessel to the containment atmosphere ensuring a low temperature overpressure event cannot jeopardize the reactor vessel integrity.

Part B There is no operational requirement or safety need for an RVV to be in the closed position when the conditions of Applicability in LCO 3.4.10 exist. However at least two valves are required to either be open or capable of being opened in response to an LTOP actuation signal when a low temperature overpressure condition could exist. A third RVV when two are not open provides allowance for the postulated single failure of one valve to open.

An RVV that is not closed has completed its safety function and providing a vent path and no further action or actuation is required. Therefore any RVV not closed is outside the scope of required components in this LCO.

An editorial correction has been made by removal of the Note at the Surveillance Requirements table. The Note was removed as unnecessary because the LCO only applies to closed reactor vent valves. A formatting error in the Bases was also corrected by removal of an extra space.

Response to Additional Discussion As noted, the NuScale LTOP function is significantly different from that of previous designs.

NuScale technical specifications are developed in close coordination and consultation with the operating staff. Experience with the technical specifications in simulator operations and in support of DCA development has not identified the need for a modified presentation of this LCO.

The staff observation of the LCO as 'unconventional' is accurate, however it is appropriate for the NuScale design. The proposed presentation would reduce clarity and introduce NuScale Nonproprietary

unnecessary complexity to the specifications. Therefore the current construction of the LCO is being retained.

Impact on DCA:

The Technical Specifications have been been revised as described in the response above and as shown in the markup provided in this response.

NuScale Nonproprietary

LTOP Valves 3.4.10 SURVEILLANCE REQUIREMENTS NOTE Not required to be met for valves that are open.

SURVEILLANCE FREQUENCY SR 3.4.10.1 Verify each RVV actuates to the open position on an In accordance with actual or simulated actuation signal. the Surveillance Frequency Control Program SR 3.4.10.2 Verify the open actuation time of each RVV is within In accordance with limits. the INSERVICE TESTING PROGRAM SR 3.4.10.3 Verify the inadvertent actuation block function for In accordance with each RVV. the Surveillance Frequency Control Program SR 3.4.10.4 Verify the inadvertent actuation block setpoint is In accordance with within limits for each RVV. the INSERVICE TESTING PROGRAM NuScale 3.4.10-2 Draft Revision 3.0

LTOP Valves B 3.4.10 BASES BACKGROUND (continued) when LTOP is required to function. Therefore the inadvertent actuation block will not prevent immediate opening of the RVVs if an LTOP actuation occurs.

With at least two RVVs open, the valves provide a vent path from the RCS to containment, preventing potential RCS low temperature overpressure conditions.

APPLICABLE Safety analyses (Ref. 3) demonstrate that the reactor vessel is SAFETY adequately protected against exceeding the Reference 1 P/T limits. In ANALYSES MODES 1 and 2, and MODE 3 with RCS cold temperature exceeding LTOP arming temperature specified in the PTLR T-1, the reactor safety valves will prevent RCS pressure from exceeding the Reference 1 limits.

Below the T-1 temperature specified in the PTLR, overpressure prevention falls to three OPERABLE or two open ECCS RVVs.

The actual temperature at which the pressure in the P/T limit curve falls below the pressurizer safety valve setpoint increases as the reactor vessel material toughness decreases due to neutron embrittlement. Each time the PTLR curves are revised, the LTOP System must be reevaluated to ensure its functional requirements can still be met using the RCS relief valve method or the depressurized and vented RCS condition.

The PTLR contains the acceptance limits that define the LTOP requirements including the setpoint for the T-1 LTOP enable interlock.

Any change to the RCS must be evaluated against the Reference 3 analyses to determine the impact of the change on the LTOP acceptance limits.

Transients that are capable of overpressurizing the RCS are categorized as either mass or heat input transients, examples of which follow:

a. Inadvertent operation of the module heatup system,

b. Excessive CVCS makeup, or

c. Spurious actuation of the pressurizer heaters.

The Reference 3 analyses demonstrate that two open RVVs can maintain RCS pressure below limits. Thus, the LCO requires each RVV to be OPERABLE or two RVVs open during the conditions when a low temperature overpressure condition could occur.

NuScale B 3.4.10-2 Draft Revision 3.0

Response to Request for Additional Information Docket No.52-048 eRAI No.: 9614 Date of RAI Issue: 10/16/2018 NRC Question No.: 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.

NuScale Nonproprietary

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.

NuScale Response:

NuScale has reviewed the proposed re-formatting and determined that the existing surveillance requirements are more appropriate for the facility design.

The wording of SR 3.7.1.1 is consistent with TSTF-GG-05-01, Rev. 1, Writers Guide for Plant-Specific Improved Technical Specifications; specifically including the discussion there of the term required as provided in section 4.1.3.

The requirements of the LCO apply to the safety and non-safetyrelated valves if they depend on an accumulator to perform their safety function. Omitting the proposed limitation ensures application of the requirement to any valve whose function depends on an accumulator.

The current construction also removes unnecessary specificity in that the term nitrogen is omitted, again resulting in the SR being applicable regardless of the pressurized accumulator design gas. For example the non-safety secondary valves could be designed to use an air-filled accumulator.

Similar points apply to the feedwater isolation and regulating valves and the construction of SR 3.7.2.1.

Impact on DCA:

There are no impacts to the DCA as a result of this response.

NuScale Nonproprietary

Response to Request for Additional Information Docket No.52-048 eRAI No.: 9614 Date of RAI Issue: 10/16/2018 NRC Question No.: 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.

NuScale Nonproprietary

NuScale Response:

Surveillance Requirement 3.6.2.5 has been deleted from the proposed generic Technical Specifications as duplicative of the requirements of technical specification 5.5.9, Containment Leakage Rate Testing Program.

Impact on DCA:

The Technical Specifications have been been revised as described in the response above and as shown in the markup provided in this response.

NuScale Nonproprietary

Containment Isolation Valves 3.6.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.1 Verify required valves accumulator pressures are In accordance with within limits. the Surveillance Frequency Control Program SR 3.6.2.2 ---------------------------------NOTE------------------------------

Valves and blind flanges in high radiation areas may be verified by use of administrative means.

Verify each containment isolation manual valve and In accordance with blind flange that is located outside containment and the Surveillance not locked, sealed, or otherwise secured and is Frequency Control required to be closed during accident conditions is Program closed, except for containment isolation valves that are open under administrative controls.

SR 3.6.2.3 Verify the isolation time of each automatic In accordance with containment isolation valve is within limits except for the INSERVICE valves that are open under administrative controls. TESTING PROGRAM SR 3.6.2.4 Verify each automatic containment isolation valve In accordance with that is not locked, sealed, or otherwise secured in the Surveillance position, actuates to the isolation position on an Frequency Control actual or simulated actuation signal except for valves Program that are open under administrative controls.

SR 3.6.2.5 Verify the combined leakage rate for all containment In accordance with bypass leakage paths is 0.6 La when pressurized the Containment to 951 psia. Leakage Rate Testing Program NuScale 3.6.2-3 Draft Revision 3.0

Containment Isolation Valves B 3.6.2 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.2.3 Verifying that the isolation time of each automatic containment isolation valve is within the limits is required to demonstrate OPERABILITY. The isolation time test ensures the valve will isolate in a time period less than or equal to that assumed in the safety analysis. Isolation time is measured from output of the module protection system equipment interface module until the valves are isolated.

An exception to the SR is provided for valves that are open under administrative control.

The isolation time and Frequency of this SR are in accordance with the INSERVICE TESTING PROGRAM.

SR 3.6.2.4 Automatic containment isolation valves close on a containment isolation signal to minimize leakage of fission products from containment and to maintain required RCS inventory following a DBA. This SR ensures each automatic containment isolation valve will actuate to its isolation position on an actual or simulated actuation signal. The Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls. An exception to the SR is also provided for valves that are open under administrative control.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.6.2.5 This SR ensures that the combined leakage rate of all containment bypass leakage paths is less than or equal to the specified leakage rate.

This provides assurance that the assumptions in the safety analysis are met. The leakage rate of each bypass leakage path is assumed to be the maximum pathway leakage (leakage through the worse of the two isolation valves) unless the penetration is isolated by use of one closed and de activated automatic valve, closed manual valve, or blind flange. In this case, the leakage rate of the isolated bypass leakage path is assumed to be the actual pathway leakage through the isolation device. If both isolation valves in the penetration are closed, the actual leakage rate NuScale B 3.6.2-7 Draft Revision 3.0

Containment Isolation Valves B 3.6.2 BASES SURVEILLANCE REQUIREMENTS (continued) is the lesser leakage rate of the two valves. The Frequency is required by the Containment Leakage Rate Testing Program. This SR simply imposes additional acceptance criteria.

Bypass leakage is considered part of La.

REFERENCES 1. 10 CFR 50, Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors, Performance-Based Requirements.

2. FSAR Section 6.2, Containment Systems.

3. FSAR Chapter 15, Transient and Accident Analysis.

NuScale B 3.6.2-8 Draft Revision 3.0

Response to Request for Additional Information Docket No.52-048 eRAI No.: 9614 Date of RAI Issue: 10/16/2018 NRC Question No.: 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.

NuScale Response:

The purpose of the surveillance frequency control program (SFCP) is to provide a means of placing surveillance test intervals under licensee control consistent with NRC policy and philosophy as expressed in numerous regulatory guides. With the test frequency under licensee control, adjustments can be made based on operating experience, test history, codes and standards, risk insights, and other factors that emerge. This approach permits adjustment of test frequency based on conditions that are not identified and information not available when the initial test frequency is established.

NuScale Nonproprietary

NuScale is the first facility to adopt the SFCP-approach to frequency management that was not previously licensed with defined frequencies in the applicable TS so FSAR Table 16.1-1 was created to describe the plant licensing bases for the initial test intervals. However the initial test intervals are only based on conditions that are expected to exist in future facilities and information available when the DCA and technical specifications were created.

The NuScale design is new and plant operations are different from historical nuclear power plants. However many initial surveillance test intervals are based on industry experience operating plants differently from NuScale procedures. Surveillance requirement (SR) 3.4.3.1 is specifically applicable during operational transient conditions as described in the associated Note. The proposed generic TS surveillance frequency for SR 3.4.3.1 is set to be in accordance with the SFCP.

NuScale proposed generic technical specification (TS) 5.5.11, Surveillance Frequency Control Program, establishes the controls that are applied to manage the surveillance frequencies and to evaluate changes when appropriate.

The initial test interval described and proposed in FSAR Table 16.1-1 for SR 3.4.3.1 is based largely on historical PWR operating practice and procedures that are different from those at a NuScale reactor. Additionally, large PWR operating experience may not be applicable to the NuScale design due to the design and operating differences between the designs.

Based on these differences between the design and operations of a historical large PWR and those of a NuScale plant the test intervals appropriate to SR 3.4.3.1 may require adjustment to ensure the limits of LCO 3.4.3, RCS P/T Limits are met.

Changes may be postulated that are required to increase, decrease, or otherwise establish more appropriate test intervals based on specific activities and changes in plant conditions that are within the scope of the surveillance tests. The application of the SFCP to this surveillance is therefore appropriate and provides a means to adjust the intervals to reflect the NuScale design, operations, and operating experience as it is accumulated.

The requirements of TS 5.5.11 are adequate to ensure adequate data and justification is developed before any changes are made. If adequate data and justification are not available, the test interval will remain at the value established in 16.1-1 consistent with that used at plants of a different design. However if data and justification becomes available the test interval may be adjusted to more appropriately reflect the NuScale power plant.

No changes to the surveillance interval for SR 3.4.3.1 are proposed.

NuScale Nonproprietary

Impact on DCA:

There are no impacts to the DCA as a result of this response.

NuScale Nonproprietary

Response to Request for Additional Information Docket No.52-048 eRAI No.: 9614 Date of RAI Issue: 10/16/2018 NRC Question No.: 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.

NuScale Response:

The requested change has been made to the description of FSAR COL Item 16.1-1 and in FSAR Table 1.8-1, Combined License Information Items.

NuScale Nonproprietary

Impact on DCA:

The Technical Specifications have been been revised as described in the response above and as shown in the markup provided in this response.

NuScale Nonproprietary

NuScale Final Safety Analysis Report Interfaces with Certified Design Table 1.8-2: Combined License Information Items (Continued)

Item No. Description of COL Information Item Section COL Item 13.6-4: A COL applicant that references the NuScale Power Plant design certification will provide 13.6 inspections, tests, analyses, and acceptance criteria for site-specific physical security structures, systems, and components (SSC).

COL Item 13.6-5: A COL applicant that references the NuScale Power Plant design certification will provide a 13.6 description of the access authorization program.

COL Item 13.6-6: A COL applicant that references the NuScale Power Plant design certification will provide a 13.6 Cyber Security Plan.

COL Item 13.7-1: A COL applicant that references the NuScale Power Plant design certification will provide a 13.7 description of the applicants 10 CFR 26 compliant fitness-for-duty (FFD) program for plant operations.

COL Item 13.7-2: A COL applicant that references the NuScale Power Plant design certification will provide a 13.7 description of the applicants 10 CFR 26 compliant fitness-for-duty (FFD) program for construction.

COL Item 14.2-1: A COL applicant that references the NuScale Power Plant design certification will describe the 14.2 site-specific organizations that manage, supervise, or execute the Initial Test Program, including the associated training requirements.

COL Item 14.2-2: A COL applicant that references the NuScale Power Plant design certification is responsible for 14.2 the development of the Startup Administration Manual that will contain the administrative procedures and requirements that control the activities associated with the Initial Test Program.

The COL applicant will provide a milestone for completing the Startup Administrative Manual and making it available for NRC inspection.

COL Item 14.2-3: A COL applicant that references the NuScale Power Plant design certification will identify the 14.2 specific operator training to be conducted during low-power testing related to the resolution of TMI Action Plan Item I.G.1, as described in NUREG-0660, NUREG-0694, and NUREG-0737.

COL Item 14.2-4: A COL applicant that references the NuScale Power Plant design certification will provide a 14.2 schedule for the Initial Test Program.

COL Item 14.2-5: A COL applicant that references the NuScale Power Plant design certification will provide a test 14.2 abstract for the potable water system pre-operational testing.

COL Item 14.2-6: A COL applicant that references the NuScale Power Plant design certification will provide a test 14.2 abstract for the seismic monitoring system pre-operational testing.

COL Item 14.2-7: A COL applicant that references the NuScale Power Plant design certification will select the plant 14.2 configuration to perform the Island Mode Test (number of NuScale Power Modules in service).

COL Item 14.3-1: A COL applicant that references the NuScale Power Plant design certification will provide the 14.3 site-specific selection methodology and inspections, tests, analyses, and acceptance criteria for emergency planning.

COL Item 14.3-2: A COL applicant that references the NuScale Power Plant design certification will provide the 14.3 site-specific selection methodology and inspections, tests, analyses, and acceptance criteria for structures, systems, and components within their scope.

COL Item 16.1-1: A COL applicant that references the NuScale Power Plant design certification will provide the 16.1 final plant-specific information identified by [ ] in the generic Technical Specifications and generic Technical Specification Bases.

COL Item 16.1-2 A COL applicant that references the NuScale Power Plant design certification will prepare and 16.1 maintain an owner-controlled requirements manual that includes owner-controlled limits and requirements described in the Bases of the Technical Specifications or as otherwise specified in the FSAR.

COL Item 17.4-1: A COL applicant that references the NuScale Power Plant design certification will describe the 17.4 reliability assurance program conducted during the operations phases of the plants life.

COL Item 17.4-2: A COL applicant that references the NuScale Power Plant design certification will identify 17.4 site-specific structures, systems, and components within the scope of the Reliability Assurance Program.

Tier 2 1.8-20 Draft Revision 2

NuScale Final Safety Analysis Report Technical Specifications Table 16.1-1 provides the initial surveillance test frequencies to be incorporated into the Surveillance Frequency Control Program (SFCP) required by NuScale GTS 5.5.11. The table identifies each GTS surveillance test requirement that references the SFCP, the base testing frequency for evaluation of future changes to the surveillance test frequency, and the basis for that test frequency.

Incorporation of Technical Specification Task Force Change Travelers Technical Specification Task Force (TSTF) travelers issued since publication of Revision 4 of the ISTS were reviewed in the development of the NuScale GTS. Travelers were incorporated into the NuScale GTS or utilized as a basis for similar NuScale situations as described in the conformance report (Reference 16.1-1). The TSTF travelers considered in development of the NuScale GTS are listed in that report.

The GTS are intended to be used as a guide in the development of the plant-specific technical specifications. Preliminary information has been provided in single brackets [ ].

Combined license applicants referencing the NuScale Power Plant are required to provide the final plant-specific information.

COL Item 16.1-1: 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.

RAI 03.06.03-11 COL Item 16.1-2: A COL applicant that references the NuScale Power Plant design certification will prepare and maintain an owner-controlled requirements manual that includes owner-controlled limits and requirements described in the Bases of the Technical Specifications or as otherwise specified in the FSAR.

16.1.2 References 16.1-1 Technical Report TTR-1116-52011, "Technical Specifications Regulatory Conformance and Development Technical Report," Rev. 0.

16.1-2 NEI 04-10, Risk-Informed Technical Specifications Initiative 5b - Risk-Informed Method for Control of Surveillance Frequencies - Industry Guidance Document, Rev. 1, April 2007.

16.1-3 NEI 06-09, Risk-Informed Technical Specifications Initiative 4b - Risk-Managed Technical Specifications (RMTS) Guidelines - Industry Guidance Document, Rev. 0-A, November 2006.

Tier 2 16.1-3 Draft Revision 2

Response to Request for Additional Information Docket No.52-048 eRAI No.: 9614 Date of RAI Issue: 10/16/2018 NRC Question No.: 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.

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 NuScale Nonproprietary

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).

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.

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.

NuScale Nonproprietary

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."

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.

NuScale Nonproprietary

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.

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.

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."

NuScale Nonproprietary

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."

NuScale Response:

Part 1 As noted in the RAI, the feedwater isolation valve (FWIV), feedwater regulating valve (FWRV),

and feedwater check valve (FCV) are described in the FSAR in Chapters 3, 5, 6, 10, and 15 because of the varied functions of the valves credited in plant analyses. The valves are credited and described to support the function of the decay heat removal system, the containment system, and support the power generation capability of the NuScale power plant.

FSAR section 3.9, Mechanical Systems and Components, describes design transients, dynamic testing and analyses, application of the ASME Code requirements, and functional design and qualification of the valves. As described in section 3.9.6, the testing requirements for valves are specified in the ASME OM Code. The NuScale inservice testing (IST) plan includes augmented testing for valves that provide a nonsafety backup of a safety-related function including establishing decay heat removal system boundary, main steam isolation and feedwater isolation.

Components subject to the IST plan are listed in FSAR table 3.9-16. The table includes the FWIV and FCV, with with details of valve testing by type provided in FSAR section 3.9.6.3.2.

Table 3.9-17 lists valves subject to augmented requirements that will be tested to the intent of the OM Code as described in FSAR 3.9.6.5.

COL Item 3.9-5 requires a COL applicant to establish an inservice testing program in accordance with ASME OM Code and 10 CFR 50.55a.

The Technical Specifications include explicit acknowledgment of, and reference to the requirements of 10 CFR 50.55a(f) in section 1.1, Definitions. Consistent with model safety evaluation for the industry / NRC traveler 545, dated December 11, 2015, the requirements of 10 CFR 50.55a will be conditions in the operating license of any nuclear reactor licensed by reference to the NuScale design certification.

NuScale Nonproprietary

Based on this, the FWIV, FWRV, and FCV are required to be tested in accordance with the IST program required by 10 CFR 50.55a, regardless of technical specification surveillance requirements. OPERABILITY of the feedwater isolation system will require that the IST program testing be accomplished in a timely manner to demonstrate the ability of the components to perform their safety or back-up credited functions.

Specific valve leakage limits for the individual valves will be developed as required by the applicable ASME Code requirements. Development of acceptance criteria will be based on the functions of the valves as described in FSAR Tables 3.9-16 and 3.9-17 including forming a portion of the decay heat removal system boundary. Specific values will be developed and implemented in the inservice testing program developed by a COL applicant as required by COL Item 3.9-5.

As described in FSAR 3.9.6.3.2, the feedwater check valve testing will be in accordance with the ASME OM Code and 10 CFR 50.55a. The basis for check valve operability is their compliance with the IST program requirements, consistent with the applicable requirements and expected condition of each facility operating license.

Part B The main steam isolation valves (MSIV) MSIV bypass valves (MSIBV), and the non-safety secondary MSIV and MSIVBV perform supporting functions for multiple credited functions including containment isolation, main steam isolation, and forming a portion of the associated decay heat removal system closed-loop boundary. The safety-related MSIV and MSIVBV are described in FSAR section 6.2.4, Containment Isolation System. The non-safety secondary MSIV and MSIVBV are described in FSAR section 10.3.2.

All eight valves (2 MSIV, 2 MSIBV, 2 non-safety secondary MSIV, and 2 non-safety secondary MSIVBV) are also addressed in FSAR section FSAR section 3.9, Mechanical Systems and Components. That section describes design transients, dynamic testing and analyses, application of the ASME Code requirements, and functional design and qualification of the valves. As described in section 3.9.6, the testing requirements for valves are specified in the ASME OM Code. The NuScale inservice testing (IST) plan includes augmented testing for valves that provide a nonsafety backup of a safety-related function including establishing decay heat removal system boundary, main steam isolation and feedwater isolation.

Components subject to the IST plan are listed in FSAR table 3.9-16. The table includes the MSIV and MSIBV, with with details of valve testing by type provided in FSAR section 3.9.6.3.2.

NuScale Nonproprietary

Table 3.9-17 lists valves subject to augmented requirements that will be tested to the intent of the OM Code as described in FSAR 3.9.6.5.

COL Item 3.9-5 requires a COL applicant to establish an inservice testing program in accordance with ASME OM Code and 10 CFR 50.55a.

The Technical Specifications include explicit acknowledgment of, and reference to the requirements of 10 CFR 50.55a(f) in section 1.1, Definitions. Consistent with model safety evaluation for the industry / NRC traveler 545, dated December 11, 2015, the requirements of 10 CFR 50.55a will be conditions in the operating license of any nuclear reactor licensed by reference to the NuScale design certification.

Based on this, the eight main steam isolation valves are required to be tested in accordance with the IST program required by 10 CFR 50.55a, regardless of technical specification surveillance requirements. OPERABILITY of the main steam isolation system will require that the IST program testing be accomplished in a timely manner to demonstrate the ability of the components to perform their safety or back-up credited functions.

Specific valve leakage limits developed for the individual valves will be developed as required by the applicable ASME Code requirements. Development of acceptance criteria will be based on the functions of the valves as described in FSAR Tables 3.9-16 and 3.9-17 including forming a portion of the decay heat removal system boundary. Specific values will be developed and implemented in the inservice testing program developed by a COL applicant as required by COL Item 3.9-5.

Impact on DCA:

There are no impacts to the DCA as a result of this response.

NuScale Nonproprietary

ML18347A619

Text

SUBJECT:

REFERENCE:

Navigation menu

Search