Text

LLC Response to NRC Request for Additional Information No. 512 (Erai No. 9634) on the NuScale Design Certification Application
	ML19029B572
Person / Time
Site:	NuScale
Issue date:	01/29/2019
From:	Rad Z; NuScale
To:	; Document Control Desk, Office of New Reactors
References
	RAIO-0119-64281
	Download: ML19029B572 (179)
	v • d • e

RAIO-0119-64281 January 29, 2019 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.

512 (eRAI No. 9634) on the NuScale Design Certification Application

REFERENCES:

1. U.S. Nuclear Regulatory Commission, "Request for Additional Information No. 512 (eRAI No. 9634)," dated November 29, 2018

2. NuScale Power, LLC Response to NRC "Request for Additional Information No. 512 (eRAI No.9634)," dated January 10, 2019

3. NuScale Power, LLC Response to NRC "Request for Additional Information No. 512 (eRAI No.9634)," dated January 16, 2019 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 responses to the following RAI Questions from NRC eRAI No. 9634:

16-60-3 16-60-26 16-60-40 16-60-7 16-60-27 16-60-42 16-60-9 16-60-28 16-60-43 16-60-17 16-60-29 16-60-46 16-60-18 16-60-30 16-60-61 16-60-21 16-60-31 16-60-62 16-60-22 16-60-32 16-60-63 16-60-23 16-60-33 16-60-72 16-60-24 16-60-34 16-60-73 16-60-25 16-60-35 16-60-78 Other portions of the NuScale response to question 16-60 were previously provided in References 2 and 3.

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

NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com

RAIO-0119-64281 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 Distribution: Gregory Cranston, NRC, OWFN-8H12 Samuel Lee, NRC, OWFN-8H12 Getachew Tesfaye, NRC, OWFN-8H12 : eRAI No. 9634, Question 16-60 Cross-Reference Table : NuScale Response to NRC Request for Additional Information eRAI No. 9634 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com

Attachment 1:

RAIO-0119-64281 eRAI No. 9634, Question 16-60 01/29/2019 Cross-Reference Table Page 1 of 3 NuScale NRC RAI Sub-Tracking paragraph NuScale Letter Submittal Letter Number Number No. Date Accession Number 16-60-1 1 RAIO-0119-64111 January 10, 2019 ML19010A409 16-60-2 2 pending 16-60-3 3 RAIO-0119-64281 January 29, 2019 pending 16-60-4 4 RAIO-0119-64111 January 10, 2019 ML19010A409 16-60-5 5 RAIO-0119-64111 January 10, 2019 ML19010A409 16-60-6 6 RAIO-0119-64178 January 16, 2019 ML19016A374 16-60-7 7 RAIO-0119-64281 January 29, 2019 pending 16-60-8 8 RAIO-0119-64111 January 10, 2019 ML19010A409 16-60-9 9 RAIO-0119-64281 January 29, 2019 pending 16-60-10 10 RAIO-0119-64111 January 10, 2019 ML19010A409 16-60-11 11 RAIO-0119-64111 January 10, 2019 ML19010A409 16-60-12 12 RAIO-0119-64111 January 10, 2019 ML19010A409 16-60-13 13 RAIO-0119-64111 January 10, 2019 ML19010A409 16-60-14 14 RAIO-0119-64111 January 10, 2019 ML19010A409 16-60-15 15 RAIO-0119-64111 January 10, 2019 ML19010A409 16-60-16 16 RAIO-0119-64111 January 10, 2019 ML19010A409 16-60-17 17 RAIO-0119-64281 January 29, 2019 pending 16-60-18 18 RAIO-0119-64281 January 29, 2019 pending 16-60-19 18 pending 16-60-20 19 RAIO-0119-64111 January 10, 2019 ML19010A409 16-60-21 20 RAIO-0119-64281 January 29, 2019 pending 16-60-22 21 RAIO-0119-64281 January 29, 2019 pending 16-60-23 22 RAIO-0119-64281 January 29, 2019 pending 16-60-24 23 RAIO-0119-64281 January 29, 2019 pending 16-60-25 24 RAIO-0119-64281 January 29, 2019 pending 16-60-26 25 RAIO-0119-64281 January 29, 2019 pending 16-60-27 26 RAIO-0119-64281 January 29, 2019 pending 16-60-28 27 RAIO-0119-64281 January 29, 2019 pending 16-60-29 28 RAIO-0119-64281 January 29, 2019 pending 16-60-30 29 RAIO-0119-64281 January 29, 2019 pending 16-60-31 29 RAIO-0119-64281 January 29, 2019 pending 16-60-32 29 RAIO-0119-64281 January 29, 2019 pending 16-60-33 30 RAIO-0119-64281 January 29, 2019 pending 16-60-34 30 RAIO-0119-64281 January 29, 2019 pending 16-60-35 31 RAIO-0119-64281 January 29, 2019 pending 16-60-36 32 RAIO-0119-64111 January 10, 2019 ML19010A409 16-60-37 33 RAIO-0119-64111 January 10, 2019 ML19010A409 16-60-38 34 RAIO-0119-64111 January 10, 2019 ML19010A409

Attachment 1:

RAIO-0119-64281 eRAI No. 9634, Question 16-60 01/29/2019 Cross-Reference Table Page 2 of 3 NuScale NRC RAI Sub-Tracking paragraph NuScale Letter Submittal Letter Number Number No. Date Accession Number 16-60-39 35 RAIO-0119-64111 January 10, 2019 ML19010A409 16-60-40 36 RAIO-0119-64281 January 29, 2019 pending 16-60-41 37.1 RAIO-0119-64111 January 10, 2019 ML19010A409 16-60-42 37.2 RAIO-0119-64281 January 29, 2019 pending 16-60-43 37.3 RAIO-0119-64281 January 29, 2019 pending 16-60-44 38 RAIO-0119-64111 January 10, 2019 ML19010A409 16-60-45 39 pending 16-60-46 40 RAIO-0119-64281 January 29, 2019 pending 16-60-47 41 RAIO-0119-64111 January 10, 2019 ML19010A409 16-60-48 42 RAIO-0119-64111 January 10, 2019 ML19010A409 16-60-49 43 RAIO-0119-64111 January 10, 2019 ML19010A409 16-60-50 44 RAIO-0119-64111 January 10, 2019 ML19010A409 16-60-51 45 RAIO-0119-64111 January 10, 2019 ML19010A409 16-60-52 45 pending 16-60-53 46 RAIO-0119-64111 January 10, 2019 ML19010A409 16-60-54 47 RAIO-0119-64111 January 10, 2019 ML19010A409 16-60-55 48 RAIO-0119-64178 January 16, 2019 ML19016A374 16-60-56 49 RAIO-0119-64111 January 10, 2019 ML19010A409 16-60-57 50 RAIO-0119-64111 January 10, 2019 ML19010A409 16-60-58 51 RAIO-0119-64111 January 10, 2019 ML19010A409 16-60-59 52 RAIO-0119-64178 January 16, 2019 ML19016A374 16-60-60 53 pending 16-60-61 54 RAIO-0119-64281 January 29, 2019 pending 16-60-62 55 RAIO-0119-64281 January 29, 2019 pending 16-60-63 55 RAIO-0119-64281 January 29, 2019 pending 16-60-64 56, 57, 58 RAIO-0119-64178 January 16, 2019 ML19016A374 16-60-65 59 pending 16-60-66 60 pending 16-60-67 61 RAIO-0119-64178 January 16, 2019 ML19016A374 16-60-68 62 RAIO-0119-64178 January 16, 2019 ML19016A374 16-60-69 63 RAIO-0119-64178 January 16, 2019 ML19016A374 16-60-70 64 RAIO-0119-64178 January 16, 2019 ML19016A374 16-60-71 65 RAIO-0119-64178 January 16, 2019 ML19016A374 16-60-72 66 RAIO-0119-64281 January 29, 2019 pending 16-60-73 67 RAIO-0119-64281 January 29, 2019 pending 16-60-74 68 pending 16-60-75 69 (i - iv) pending 16-60-76 69 (v) pending 16-60-77 70 pending

Attachment 1:

RAIO-0119-64281 eRAI No. 9634, Question 16-60 01/29/2019 Cross-Reference Table Page 3 of 3 NuScale NRC RAI Sub-Tracking paragraph NuScale Letter Submittal Letter Number Number No. Date Accession Number 16-60-78 71 RAIO-0119-64281 January 29, 2019 pending 16-60-79 72 pending 16-60-80 73 pending 16-60-81 74 RAIO-0119-64111 January 10, 2019 ML19010A409 16-60-82 75 pending

RAIO-0119-64281 :

NuScale Response to NRC Request for Additional Information eRAI No. 9634 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.: 9634 Date of RAI Issue: 11/29/2018 NRC Question No.: 16-60-3

3. A response is not required because this item is being referred to the NRO technical branch responsible for reviewing FSAR Section 6.4. Markup of Page 6.4-1 of FSAR Section 6.4 in Letter dated June 1, 2018, follow-up to NRC-NuScale public meetings held on 2/26/2018 and 4/3/2018. (Also see response to RAI 01-1.) NuScale is requested to explain why it proposed adding the sentence, No operator actions are required or credited to mitigate the consequences of design basis events, before or after 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . This sentence is included in Revision 2 of DCA part 2, FSAR Section 6.4.

NuScale Response:

FSAR Tier 2, section 15.0.0.6.4, Required Operator Actions states:

There are no operator actions credited in the evaluation of NuScale DBEs. After a DBE, automated actions place the NPM in a safe-state and it remains in the safe-state condition for at least 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months without operator action, even with assumed failures.

This is a feature of the NuScale power plant design that is reflected throughout the design and analyses presented in the FSAR including Tier 2, chapters 6, 9, and 15.

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.: 9634 Date of RAI Issue: 11/29/2018 NRC Question No.: 16-60-7

7. Please coordinate the response to this sub-question with the response to similar Sub-question 37.3. The applicant is requested to explain whether the N-2L- Power Range Linear Power Permissive, which allows manually bypassing of MPS Functions 3.3.1.1.a and 3.3.1.1.b with Thermal Power > 15% RTP, should be referred to as the N-2L interlock when it automatically enables these Functions at 15% RTP. The staff needs to ensure that the terms

'permissive' and 'interlock' are used consistently when referring to MPS operating bypass and enable Functions.

NuScale Response:

Module Protection System interlocks and permissives, including the N-2L Permissive, N-2L Interlock, and N-2H Interlock are described in FSAR Tier 2, Table 7.1-5. Interlocks automatically establish an operating bypass when the interlock condition is met. Permissives allow the manual bypass by the operator of the operating bypass. Operating bypasses are automatically removed when the associated interlock condition or permissive condition is no longer satisfied.

This information is also described in the Applicable Safety Analyses, LCO, and Applicability section of the bases for LCO 3.3.1. As described in the FSAR and the bases, the distinction between functions as used in the NuScale design is that a permissive feature allows operator action to implement the bypass, while interlock features automatically implement the bypass.

Both permissive and interlock functions are automatically removed when the associated condition no longer exists.

NuScale Nonproprietary

The N-2 permissive and interlocks are distinguished by the appended L or H which indicates whether the condition is entered from below or above the setpoint. N-2L functions are active on increasing power. The N-2H function is active on decreasing power.

MPS function 3.3.1.1.a is described in the bases correctly as a permissive. The permissive may be initiated by the operator when power exceeds the N-2L setpoint from below. The automatic reset as power decreases below the permissive setpoint is consistent with the design and descriptions used for permissive functions.

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.: 9634 Date of RAI Issue: 11/29/2018 NRC Question No.: 16-60-9

9. A separate response is not required because the response to this sub-question will be provided with a supplemental response to RAI 9034, Question 16-30, Sub-question a.1. The proposed definition of SDM in Revision 2 of DCA part 4, Section 1.1, departs from the W-STS definition as indicated by the following mark up of the W-STS definition:

SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:

a. Moderator temperature is 420°F; and

b. All rod cluster control assemblies (RCCAs) CRAs are fully inserted except for the single RCCA assembly of highest reactivity worth, which is assumed to be fully withdrawn. However, with all RCCAs CRAs verified fully inserted by two independent means, it is not necessary to account for a stuck RCCA CRA in the SDM calculation. With any RCCA CRA not capable of being fully inserted, the reactivity worth of the RCCA the affected CRA must be accounted for in the determination of SDM, and

b. In MODES 1 and 2, the fuel and moderator temperatures are changed to the [nominal zero power design level].

The change in the order of parts a and b, and the use of CRA instead of RCCA, are editorial administrative changes to reflect NuScale nomenclature and the applicant's preferred presentation. Since the acronym "CRA" is previously defined in the definition of "MODE," not defining it upon its first use in this definition, is acceptable. However, the staff suggests defining the acronym again for clarity. Regardless, subsequent use of the word "assembly" and "assemblies" should be changed to "CRA" and "CRAs" to conform to the improved TS writer's NuScale Nonproprietary

guide convention concerning acronyms. The W-STS definition does not appear to consider more than one RCCA to be incapable of being fully inserted; however, the W-AP1000-STS SDM definition does consider more than one uninsertable RCCA. Revision 1 of the DCA contains no justification of why NuScale needs to consider more than one CRA that cannot be fully inserted.

Finally, the DCA does not justify using the minimum temperature for criticality, 420°F, in place of the statement, "In MODE 1, the fuel and moderator temperatures are changed to the [nominal zero power design level]." (Note that NuScale MODE 1 corresponds to W-STS MODES 1 and 2; and NuScale MODE 2 corresponds to W-STS MODE 3 with RCS average temperature 420°F.) The applicant is requested to resolve these issues by providing the noted missing justifications, provided they are acceptable to the staff, and by editing the SDM definition to state:

SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:

a. Moderator temperature is 420°F; and

b. All control rod assemblies (CRAs) are fully inserted except for the single CRA of highest reactivity worth, which is assumed to be fully withdrawn. However, with all CRAs verified fully inserted by two independent means, it is not necessary to account for a stuck CRA in the SDM calculation. With any CRA not capable of being fully inserted, the reactivity worth of the affected CRA must be accounted for in the determination of SDM.

NuScale Response:

The acronym CRA was re-defined in the definition of SHUTDOWN MARGIN (SDM), and the acronym CRA is used in place of the term assembly as requested. Changes were also made to more closely align with the Westinghouse STS definition.

The choice of 420 °F as the reference temperature for shutdown margin calculation is a design choice; regardless of the reference temperature value chosen, the primary importance is to assure consistent calculation of SDM. 420 °F was selected based on the minimum temperature for criticality established by LCO 3.4.2 and it being the limiting zero power design temperature.

The temperature is also an appropriate reference point for use in NuScale procedures and analyses. For example, core operating limits report limits related to SDM will be established by reference to this selected temperature. Another example is the use of the 420 °F reference temperature in FSAR Tier 2, Section 15.0.6.3.2, Input Parameters and Initial Conditions.

NuScale Nonproprietary

The NuScale plant operates in a manner that varies reactor average reactor coolant temperature from approximately 425 °F to 545 °F as reactor power increases from 0% rated thermal power (RTP) to 15% RTP, and remains constant above 15% RTP. The use of 420 °F somewhat simplifies operations and procedures used to calculate SDM. These operations are described in Table 5.1-2 of the FSAR, Tier 2 description of the RCS and connecting systems.

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

Definitions 1.1 1.1 Definitions SHUTDOWN MARGIN (SDM) SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:

a. Moderator temperature is 420 °F; and

b. All control rod assemblies (CRAs) are fully inserted except for the single CRAassembly of highest reactivity worth, which is assumed to be fully withdrawn.

However, with all CRAs verified fully inserted by two independent means, it is not necessary to account for a stuck CRA in the SDM calculation. With any CRA(s) not capable of being fully inserted, the reactivity worth of these affected CRAassemblies must be accounted for in the determination of SDM.

THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

NuScale 1.1-6 Draft Revision 3.0

Response to Request for Additional Information Docket No.52-048 eRAI No.: 9634 Date of RAI Issue: 11/29/2018 NRC Question No.: 16-60-17

17. A response is not required because in the November 6, 2018, public meeting conference call, the applicant stated the requested changes will be incorporated in Revision 3 of DCA part

2. In Revision 2 of DCA part 2, FSAR Tier 2 page 3.9-40, last paragraph of Subsection 3.9.4.1.1 under heading Sensor Coil Assembly. The first sentence should be two sentences with the indicated corrections:

The sensor coil assembly contains the rod position indication coils. The coil coils the coil assembly slides over the rod travel housing and sits sets on a ledge at the base of the rod travel housing.

NuScale Response:

The location of the sensor coil assembly as been clarified in FSAR 3.9.4.1.1.

Impact on DCA:

The FSAR has 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 Mechanical Systems and Components position. The stepping process now only moves the outer shaft. The control rod drive shaft is lowered, which retracts the plug on the center disconnect rod from the fingers on the coupling, and inserts the fingers into the CRA hub. The additional step compresses the spring in the CRA hub slightly, and ensures the fingers on the coupling are completely seated. The plug on the bottom of the center disconnect rod is inserted by releasing the remote disconnect gripper. The center disconnect rod then falls, with spring assist, to lock the control rod drive shaft to the CRA hub.

A lift verification is then performed by observing the difference in lift coil current confirming successful completion of the remote reconnect operation.

RAI 03.09.04-1, RAI 03.09.04-1S1, RAI 03.09.04-2, RAI 03.09.04-2S1, RAI 03.09.04-4, RAI 03.09.04-4S1, RAI 03.09.04-5, RAI 03.09.04-6, RAI 03.09.04-7, RAI 03.09.04-9 In the event that the control rod drive shaft cannot be remotely disconnected from the CRA remotely, an alternate non-remote method is provided to disengage the CRA through the top of the rod travel housing (Figure 4.6-4). Since operation of the remote disconnect mechanism requires the entire CRDM to be operational, there are a number of reasons that could prevent an intentional remote disconnect. This includes, but is not limited to, the inability of the stationary gripper or remote disconnect gripper latches to properly engage, either due to a mechanical failure of the latches, a failure of the drive coils, or a failure of the disconnect verification. In the event that the remote disconnect mechanism operation is not available, the pressure boundary seal weld around the rod travel housing plug is broken, and the plug is removed for tooling access. The top of the control rod drive shaft contains a locking feature that allows for manual lift of the remote disconnect rod and unlock the CRA (Figure 4.6-6).

RAI 03.09.04-1, RAI 03.09.04-2, RAI 03.09.04-4, RAI 03.09.04-5, RAI 03.09.04-7, RAI 03.09.04-9 Drive Coil Assembly The drive coil assembly slides over the latch housing and sets on a ledge at the base of the latch housing. The drive coil assembly is depicted by Figure 4.6-3.

Sensor Coil Assembly RAI 03.09.04-1, RAI 03.09.04-2, RAI 03.09.04-4, RAI 03.09.04-5, RAI 03.09.04-6, RAI 03.09.04-7, RAI 03.09.04-9, RAI 16-60 The sensor coil assembly contains the rod position indication coils. tThe coil assembly slides over the rod travel housing and siets on a ledge at the base of the rod travel housing. The sensor coil assembly is shown in Figure 4.6-4.

3.9.4.1.2 Operation of the Control Rod Drive Mechanisms RAI 03.09.04-1, RAI 03.09.04-2, RAI 03.09.04-4, RAI 03.09.04-5, RAI 03.09.04-6, RAI 03.09.04-7, RAI 03.09.04-9 The basic CRDM mechanical and operational requirements are discussed in Section 4.6. The following describes the different modes of CRDM operation.

Reactor trip, consisting of full insertion of the CRAs into the core at design conditions, is achievable during any part of the CRDM operating modes described below.

Tier 2 3.9-42 Draft Revision 3

Response to Request for Additional Information Docket No.52-048 eRAI No.: 9634 Date of RAI Issue: 11/29/2018 NRC Question No.: 16-60-18

18. The applicant is requested to make the indicated changes or make appropriate equivalent changes:

18.1 A response is not required because in the November 6, 2018, public meeting conference call, the applicant stated the requested change will be incorporated in Revision 3 of DCA part 4.

The applicant is requested to change Revision 2 of DCA part 4, Condition A of Subsection 3.1.5 to be consistent with Condition A of Subsection 3.1.6, which is more consistent with STS style and phrasing conventions, as follows: "Shutdown group not within insertion limits not met."

NuScale Response:

See the response to RAI 16-60-26 which aligns the Action requirements and Bases of LCO 3.1.5 and 3.1.6.

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.: 9634 Date of RAI Issue: 11/29/2018 NRC Question No.: 16-60-21

20. In Revision 2 of DCA part 4, GTS Bases page B 3.1.5-1, Background section:

20.1 The applicant is requested to clarify the first sentence as indicated:

The insertion limits of the shutdown bank group control rod assemblies (CRAs) are initial assumptions in all safety analyses that assume shutdown group bank CRA insertion upon reactor trip.

The revised sentence is consistent with the design terminology ("shutdown bank CRAs"),

which is used in the first sentence of the third paragraph of the Background section. Also, there is no need to capitalize "control rod assemblies."

20.2 The applicant is requested to clarify the third paragraph as indicated:

The shutdown bank CRAs are arranged into two groups; each group has four CRAs that are arranged in radially symmetric positions, and are normally moved together as a group.

Therefore, movement of the shutdown a group of shutdown bank CRAs does not introduce radial asymmetries in the core power distribution. The shutdown bank and regulating group bank CRAs provide the required reactivity worth for immediate reactor shutdown upon a reactor trip.

20.3 The applicant is requested to clarify the fourth paragraph as indicated:

The design calculations are performed with the assumption that the CRA shutdown group CRAs are withdrawn prior to the CRA regulating group CRAs. The CRA shutdown group CRAs can be fully withdrawn without the core going critical. This provides available negative reactivity for SDM in the event of unintended dilution of the RCS boron concentration. The CRA shutdown group CRAs are controlled manually or automatically by the control room operator. During NuScale Nonproprietary

normal unit operation, the CRA shutdown group CRAs are fully withdrawn. The CRA shutdown group CRAs must be completely withdrawn from the core prior to withdrawing the CRA regulating group CRAs during an approach to criticality. The CRA shutdown group CRAs are then left in this the fully withdrawn position until the reactor is shut down. They The eight CRAs of the shutdown bank add negative reactivity to shut down the reactor upon receipt of a reactor trip signal The applicant is also requested to insert in the above paragraph (1) an explanation of how automatic control "by the control room operator" is distinct from manual control "by the control room operator"; and (2) a statement about when automatic control of [movement of] CRA shutdown groups is appropriate and designed to be used.

NuScale Response:

The bases have been modified.

A description of the control rod drive system and its operation is provided in FSAR Tier 2, section 4.6. FSAR Tier 2, section 7.0.4.5 describes the Module Control System (MCS), including the control rod drive system control inputs.

Detailed system design has not been completed, and operating procedures have not been finalized. Procedure finalization will be completed as described in COL Item 13.5-2. The procedures will describe the use of the manual and automatic control alternatives consistent with the licensing basis description in the FSAR. For example, FSAR Tier 2, section 4.1.1 describes the means of reactivity control and arrangement of the CRA in the reactor.

As noted in FSAR section 7.0.4.5.1, rod withdrawal is performed by an operator, although supervisory control of rod insertion may be performed by the MCS controller if reactor power demand requires rod insertion to reduce power. Automatic rod withdrawal may not be performed when thermal reactor power is between zero and 15 percent.

The safety function of the module protection system (MPS) and control rod drive system is independent of the MCS and control rod drive system control portion of the system. If a reactor trip signal is received, the MPS interrupts the electrical power to the control rod drive system de-energizing the mechanism holding the control rods out of the core. This causes the rods to insert by gravity.

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

Shutdown GroupBank Insertion Limits 3.1.5 3.1 REACTIVITY CONTROL SYSTEMS 3.1.5 Shutdown GroupBank Insertion Limits LCO 3.1.5 Each shutdown bank group shall be within insertion limits specified in the COLR.

NOTE Not applicable to shutdown groups inserted while performing SR 3.1.4.2.

APPLICABILITY: MODE 1.

NOTE---------------------------------------------

This LCO not applicable while performing SR 3.1.4.2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Shutdown group not A.1.1 Verify SDM is within the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months within limitsOne or limits specified in the more shutdown groups COLR.

not within insertion limits. OR A.1.2 Initiate boration to restore 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months SDM to within limit.

AND A.2 Restore shutdown groups 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to within limits.

B. Required Action and B.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.

NuScale 3.1.5-1 Draft Revision 3.0

Shutdown GroupBank Insertion Limits 3.1.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.5.1 Verify each shutdown bank group is within the In accordance insertion limits specified in the COLR. with the Surveillance Frequency Control Program NuScale 3.1.5-2 Draft Revision 3.0

Regulating GroupBank Insertion Limits 3.1.6 3.1 REACTIVITY CONTROL SYSTEMS 3.1.6 Regulating GroupBank Insertion Limits LCO 3.1.6 Each regulating bank group shall be within the insertion limits specified in the COLR.

NOTE Not applicable to regulating groups inserted while performing SR 3.1.4.2.

APPLICABILITY: MODE 1 with keff 1.0.

NOTE---------------------------------------------

This LCO not applicable while performing SR 3.1.4.2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more regulating A.1.1 Verify SDM is within the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months groups not within limits specified in the insertion COLR.

limits.Regulating group insertion limits not met. OR A.1.2 Initiate boration to restore 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months SDM to within limits.

AND A.2 Restore regulating groups 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to within limits.

B. Required Action and B.1 Be in MODE 1 with 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion keff < 1.0.

Time not met.

NuScale 3.1.6-1 Draft Revision 3.0

Regulating GroupBank Insertion Limits 3.1.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.6.1 Verify each regulating bank group is within the In accordance with insertion limits specified in the COLR. the Surveillance Frequency Control Program NuScale 3.1.6-2 Draft Revision 3.0

PHYSICS TESTS Exceptions 3.1.8 3.1 REACTIVITY CONTROL SYSTEMS 3.1.8 PHYSICS TESTS Exceptions LCO 3.1.8 During the performance of PHYSICS TESTS, the requirements of:

LCO 3.1.3, Moderator Temperature Coefficient (MTC),

LCO 3.1.4, Rod Group Alignment Limits, LCO 3.1.5, Shutdown BankGroup Insertion Limits, and LCO 3.1.6, Regulating BankGroup Insertion Limits may be suspended provided:

a. SDM is within the limits specified in the COLR, and

b. THERMAL POWER is 5% RTP.

APPLICABILITY: During PHYSICS TESTS initiated in MODE 1.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. SDM not within limit. A.1 Initiate boration to restore 15 minutes SDM to within limit.

AND A.2 Suspend PHYSICS TESTS 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months exceptions.

B. THERMAL POWER not B.1 Open reactor trip breakers. Immediately within limit.

NuScale 3.1.8-1 Draft Revision 3.0

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.3 Core Operating Limits Report (COLR)

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

3.1.1, SHUTDOWN MARGIN (SDM);

3.1.3, Moderator Temperature Coefficient (MTC);

3.1.4, Rod Group Alignment Limits; 3.1.5, Shutdown BankGroup Insertion Limits; 3.1.6, Regulating BankGroup Insertion Limits; 3.1.8, PHYSICS TESTS Exceptions; 3.1.9, Boron Dilution Control; 3.2.1, Enthalpy Rise Hot Channel Factor ( F H);

3.2.2, AXIAL OFFSET (AO);

3.4.1, RCS Pressure, Temperature, and Flow Resistance Critical Heat Flux (CHF) Limits; and 3.5.3, Ultimate Heat Sink.

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

[-----------------------------------REVIEWER'S NOTE----------------------------------

The COL applicant shall confirm the validity of each listed document and the listed Specifications for the associated core operating limits, or state the valid NRC approved analytical method document and list of associated Specifications.

The COL applicant shall state the valid core reload analysis methodology document and list of associated Specifications.

]

NuScale 5.6-2 Draft Revision 3.0

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.3 Core Operating Limits Report (COLR) (continued)

1. [NuScale Standard Design Certification Analysis (DCA), Part 2, Tier 2, NuScale Final Safety Analysis Report (FSAR), Section 4.3, Nuclear Design, Revision 1, March 2018, and TR-0516-49416, Non-Loss-of-Coolant Accident Analysis Methodology, Revision 0, May 2016 (NuScale Proprietary).

(Methodology for Specifications 3.1.1 - SHUTDOWN MARGIN (SDM), 3.1.3 - Moderator Temperature Coefficient, 3.1.4 - Rod Group Alignment Limits, 3.1.5 - Shutdown BankGroup Insertion Limits, 3.1.6 - Regulating BankGroup Insertion Limits, and 3.1.8 -

PHYSICS TESTS Exceptions.)]

2. [NuScale DCA, Part 2, Tier 2, NuScale FSAR, Section 9.3.4, Chemical and Volume Control System, Revision 1, March 2018, and TR-0516-49416, Non-Loss-of-Coolant Accident Analysis Methodology, Revision 0, May 2016 (NuScale Proprietary).

(Methodology for Specification 3.1.9 - Boron Dilution Control.)]

3. [NuScale DCA, Part 2, Tier 2, NuScale FSAR, Sections 4.3, Nuclear Design, and 4.4, Thermal and Hydraulic Design, Revision 1, March 2018; TR-0516-49416, Non-Loss-of-Coolant Accident Analysis Methodology, Revision 0, May 2016 (NuScale Proprietary); and TR-0915-17564, Subchannel Analysis Methodology, Revision 1, September 2015 (NuScale Proprietary).

(Methodology for Specifications 3.2.1 - Enthalpy Rise Hot Channel Factor (FH), and 3.2.2 - AXIAL OFFSET (AO).)]

4. [NuScale DCA, Part 2, Tier 2, NuScale FSAR, Section 4.4, Thermal and Hydraulic Design, Revision 1, March 2018 and TR-0516-49416, Non-Loss-of-Coolant Accident Analysis Methodology, Revision 0, May 2016 (NuScale Proprietary).

(Methodology for Specification 3.4.1 - RCS Pressure, Temperature, and Flow Resistance CHF Limits.)]

5. [NuScale DCA, Part 2, Tier 2, NuScale FSAR, Section 4.3, Nuclear Design, Revision 1, March 2018.

(Methodology for Specifications 3.5.3 - Ultimate Heat Sink, and 3.8.1 - Nuclear Instrumentation.)]

NuScale 5.6-3 Draft Revision 3.0

SDM B 3.1.1 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.1 SHUTDOWN MARGIN (SDM)

BASES BACKGROUND According to GDC 26 (Ref. 1) the reactivity control systems must be redundant and capable of holding the reactor core subcritical when shutdown under cold conditions. Maintenance of the SDM ensures that postulated reactivity events will not damage the fuel.

SDM requirements provide sufficient reactivity margin to assure that specified acceptable fuel design limits (SAFDLs) will not be exceeded for normal shutdown and anticipated operational occurrences (AOOs).

As such, the SDM defines the degree of subcriticality that would be obtained immediately following the insertion or scram of all shutdown and regulating bankgroup control rod assemblies (CRAs), assuming that the single CRA of highest reactivity worth is fully withdrawn.

Additionally SDM requirements provide sufficient reactivity margin to ensure that the reactor will remain shutdown at all temperatures with all control rods inserted.

The system design requires that two independent reactivity control systems be provided, and that one of these systems be capable of maintaining the core subcritical under cold conditions. These requirements are provided by the use of movable CRAs and soluble boric acid in the Reactor Coolant System (RCS). The CRA System provides the SDM during power operation and is capable of making the core subcritical rapidly enough to prevent exceeding acceptable fuel damage limits, following all AOOs and postulated accidents, assuming that the CRA of highest reactivity worth remains withdrawn.

The soluble boron system can compensate for fuel depletion during operation and all xenon burnout reactivity changes and maintain the reactor subcritical under cold conditions.

During power operation, SDM control is ensured by operating with the shutdown bank groups fully withdrawn and the regulating bank groups within the limits of LCO 3.1.6, Regulating BankGroup Insertion Limits.

When the unit is in MODES 2, 3, 4 or 5, the SDM requirements are met by means of adjustments to the RCS boron concentration and the boron requirements for the pool, LCO 3.5.3, "Ultimate Heat Sink" and CRA controls.

NuScale B 3.1.1-1 Draft Revision 3.0

SDM B 3.1.1 BASES APPLICABLE SAFETY ANALYSES (continued)

SDM satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). Even though it is not directly observed from the main control room, SDM is considered an initial condition process variable because it is periodically monitored to ensure that the unit is operating within the bounds of accident analysis assumptions.

LCO SDM is a core design condition that can be ensured during operation through CRA positioning (regulating and shutdown groupbanks) and through the soluble boron concentration.

APPLICABILITY In MODE 1 with keff 1.0, SDM requirements are ensured by complying with LCO 3.1.5, "Shutdown BankGroup Insertion Limits,"

and LCO 3.1.6, "Regulating BankGroup Insertion Limits."

In MODE 1 with keff < 1.0 and in MODES 2, 3, and 4, the SDM requirements are applicable to provide sufficient negative reactivity to meet the assumptions of the safety analyses discussed above.

In MODE 5 the shutdown reactivity requirements are given in LCO 3.5.3, "Ultimate Heat Sink.

ACTIONS A.1 If the SDM requirements are not met, boration must be initiated promptly. A Completion Time of 15 minutes is adequate for an operator to correctly align and start the required systems and components. It is assumed that boration will be continued until the SDM requirements are met.

In the determination of the required combination of boration flow rate and boron concentration, there is no unique requirement that must be satisfied. Since it is imperative to raise the boron concentration of the RCS as soon as possible, the boron concentration should be a concentrated solution. The operator should begin boration with the best source available for the plant conditions.

NuScale B 3.1.1-3 Draft Revision 3.0

Rod Group Alignment Limits B 3.1.4 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.4 Rod Group Alignment Limits BASES BACKGROUND The OPERABILITY (i.e., trippability) of the shutdown and regulating control rod assemblies (CRAs) is an initial assumption in all safety analyses that assume CRA insertion upon reactor trip. Maximum CRA misalignment is an initial assumption in the safety analysis that directly affects core power distributions and assumptions of available shutdown margin (SDM).

The applicable criteria for these reactivity and power distribution design requirements are 10 CFR 50, Appendix A, GDC 10, Reactor Design, and GDC 26, Reactivity Control System Redundancy and Capability (Ref. 1), and 10 CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Plants (Ref. 2).

Mechanical or electrical failures may cause a CRA to become inoperable or to become misaligned from its group. CRA inoperability or misalignment may cause increased power peaking, due to the asymmetric reactivity distribution and a reduction in the total available CRA worth for reactor shutdown. Therefore, CRA alignment and OPERABILITY are related to core operation in design power peaking limits and the core design requirement of a minimum SDM.

Sixteen CRAs are arranged in four symmetrical groups. There are two shutdown bank groups of four CRAs each and two regulating bank groups of four CRAs each.

Limits on CRA alignment and OPERABILITY have been established, and CRA positions are monitored and controlled during power operation to ensure that the power distribution and reactivity limits defined by the design power peaking and SDM limits are preserved.

CRAs are moved by their control rod drive mechanisms (CRDMs).

Each CRDM moves its CRA one step (approximately 3/8 inch) at a time.

The CRAs are arranged into groups that are radially symmetric.

Therefore, movement of the CRAs by group does not introduce radial asymmetries in the core power distribution. The shutdown and regulating CRAs provide the required reactivity worth for immediate reactor shutdown upon a reactor trip. The regulating bank CRAs also provide power level control during normal operation and transients.

NuScale B 3.1.4-1 Draft Revision 3.0

Rod Group Alignment Limits B 3.1.4 BASES BACKGROUND (continued)

Their movement may be automatically controlled by the reactivity control systems.

The axial position of shutdown and regulating group CRAs is indicated by two separate and independent rod position indication systems.

APPLICABLE CRA misalignment accidents are analyzed in the safety analysis SAFETY (Ref. 3). The accident analysis defines CRA misoperation as any event ANALYSES with the single failure of a safety-related component and multiple failures of non-safety related controls. The acceptance criteria for addressing CRA inoperability or misalignment are that:

a. With the most reactive CRA stuck out of the core there will be no violations of either:

1. Specified acceptable fuel design limits (SAFDLs); or

2. Reactor Coolant System (RCS) pressure boundary integrity; and

b. The core must remain subcritical after design basis events with all CRAs fully inserted.

Accident and transient analyses associated with CRA misalignment, static and dynamic, account for misalignment of 6 steps at the initiation of the event. The results of the CRA misoperation analysis show that during the most limiting misoperation events, no violations of the SAFDLs, or the SLs on critical heat flux ratio, fuel centerline temperature, or pressurizer pressure occur.

CRA alignment limits and OPERABILITY requirements satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii).

NuScale B 3.1.4-2 Draft Revision 3.0

Rod Group Alignment Limits B 3.1.4 BASES ACTIONS (continued) significant, and operation may proceed without further restriction. An alternative to realigning a single misaligned CRA to the group average position is to align the remainder of the group to the position of the misaligned CRA. However, this must be done without violating the groupbank sequence, overlap, and insertion limits specified in LCO 3.1.5, "Shutdown BankGroup Insertion Limits," and LCO 3.1.6, "Regulating BankGroup Insertion Limits." The Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is adequate for determining SDM and, if necessary, for initiating boration and restoring SDM.

In this situation, SDM verification must include the worth of any untrippable CRA, in addition to the CRA of maximum worth.

A.2 When Required Action cannot be completed within their Completion Time, the unit must be brought to a MODE or Condition in which the LCO requirements are not applicable. To achieve this status, the unit must be brought to at least MODE 2 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , which obviates concerns about the development of undesirable xenon and power distributions. The allowed Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable, based on operating experience, for reaching MODE 2 from full power conditions in an orderly manner.

SURVEILLANCE SR 3.1.4.1 REQUIREMENTS Verification that the position of individual rods is within alignment limits allows the operator to detect that a rod is beginning to deviate from its expected position.

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

The SR is modified by a Note that permits it not to be performed for rods associated with an inoperable rod position indicator. The alignment limit is based on rod position indicator which is not available if the indicator is inoperable. LCO 3.1.7, Rod Position Indication, provides Actions to verify the rods are in alignment when one or more rod position indicators are inoperable.

NuScale B 3.1.4-4 Draft Revision 3.0

Shutdown GroupBank Insertion Limits B 3.1.5 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.5 Shutdown BankGroup Insertion Limits BASES BACKGROUND The insertion limits of the shutdown group bank Control control Rod rod Assemblies assemblies (CRAs) are initial assumptions in all safety analyses that assume shutdown group bank CRA insertion upon reactor trip. The insertion limits directly affect core power distributions and assumptions of available shutdown margin (SDM), ejected CRA worth, and initial reactivity insertion rate.

The applicable criteria for these reactivity and power distribution design requirements are 10 CFR 50, Appendix A, GDC 10, "Reactor Design," and GDC 26, "Reactivity Limits" (Ref. 1), and 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors" (Ref. 2). Limits on shutdown bankgroup CRA insertion have been established, and all shutdown bankgroup CRA positions are monitored and controlled during power operation to ensure that the reactivity limits, ejected CRA worth, and SDM limits are preserved.

The 16 CRAs are divided among the two regulating bank groups and two shutdown bank groups, with each group consisting of four CRAs in radially symmetric core locations. The shutdown bank CRAs are arranged into groups that are radially symmetricnormally moved together as a group. Therefore, movement of the shutdowna group of shutdown bank CRAs does not introduce radial asymmetries in the core power distribution. The shutdown bank and regulating group bank CRAs provide the required reactivity worth for immediate reactor shutdown upon a reactor trip.

The design calculations are performed with the assumption that CRAs of the shutdown group CRAsbank are withdrawn prior to the CRAs in the regulating group bankCRAs. The CRAs of the shutdown group CRAsbank can be fully withdrawn without the core going critical. This provides available negative reactivity for SDM in the event of unintended reduction of the RCS boron concentrationdilution. The shutdown groupbank CRAs are controlled manually or automatically by the control room operator. During normal unit operation, the shutdown groupbank CRAs are fully withdrawn. The shutdown groupbank CRAs must be completely withdrawn from the core prior to withdrawing regulating groupbank CRAs during an approach to criticality. The shutdown NuScale B 3.1.5-1 Draft Revision 3.0

Shutdown GroupBank Insertion Limits B 3.1.5 BASES BACKGROUND (continued) groupbank CRAs are then left in this the fully withdrawn position until the reactor is shut down. They eight CRAs of the shutdown bank add negative reactivity to shut down the reactor upon receipt of a reactor trip signal.

APPLICABLE On a reactor trip, all CRAs (shutdown groupeight CRAs in two SAFETY shutdown bank groups and eight CRAs in two regulating bank group ANALYSES CRAs), except the most reactive CRA, are assumed to insert into the core. The shutdown groupbank and regulating groupbank CRAs shall be at or above their insertion limits and available to insert the maximum amount of negative reactivity on a reactor trip signal. The regulating groupbank CRAs may be partially inserted in the core as allowed by LCO 3.1.6, "Regulating GroupBank Insertion Limits." The shutdown and regulating groupbank CRA insertion limits are established to ensure that a sufficient amount of negative reactivity is available to shut down the reactor and maintain the required SDM (see LCO 3.1.1, "SHUTDOWN MARGIN (SDM)") following a reactor trip from full power. The combination of regulating group CRAs and shutdown groupbank CRAs (less the most reactive CRA, which is assumed to be fully withdrawn) isare sufficient to take the reactor from full power conditions at rated temperature to zero power, and to maintain the required SDM at rated no load temperature (Ref. 3). The CRA shutdown groupbank CRA insertion limits also limits ensure that the reactivity worth of an ejected shutdown CRA is within safety analysis assumptions.

The acceptance criteria for addressing CRA shutdown CRA bank and as well as regulating groupbank CRA insertion limits and CRA inoperability or misalignment are that:

a. With the most reactive CRA stuck out there will be no violation of either:

1. Specified acceptable fuel design limits; or

2. Reactor Coolant System pressure boundary damage integrity; and

b. The core remains subcritical after design basis events with all CRAs fully inserted.

The CRA shutdown groupbank CRA insertion limits satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii).

NuScale B 3.1.5-2 Draft Revision 3.0

Shutdown GroupBank Insertion Limits B 3.1.5 BASES LCO The CRA shutdown CRAs bank must be within their insertion limits any time the reactor is critical or approaching criticality. This ensures that a sufficient amount of negative reactivity is available to shut down the reactor and maintain the required SDM following a reactor trip.

The CRA shutdown groupbank insertion limits are defined specified in the COLR.

The LCO is modified by a Note indicating the LCO requirement is not applicable to shutdown groups being inserted while performing SR 3.1.4.2. The SR verifies the freedom of the rods to move, and may require the shutdown group to move below the LCO limits, which normally violate the LCO. This Note applies to each shutdown group as its moved below the insertion limit to perform the SR. This Note is not applicable should a malfunction stop performance of the SR.

APPLICABILITY The CRA shutdown groupbank CRAs must be within their insertion limits, with the reactor in MODE 1. This ensures that a sufficient amount of negative reactivity is available to shut down the reactor and maintain the required SDM following a reactor trip. In MODE 2, 3, 4 the shutdown groupbank CRAs are fully inserted in the core and contribute to the SDM. Refer to LCO 3.1.1, "SHUTDOWN MARGIN (SDM)," for SDM requirements in MODES 2, 3, and 4. LCO 3.5.3, "Ultimate Heat Sink," ensures adequate SDM in MODES 4 and 5.

The Applicability is modified by a Note indicating the LCO requirement is not applicable while performing SR 3.1.4.2. This Note permits exceeding the CRA shutdown bank insertion limits while inserting each CRA in the bank in accordance with SR 3.1.4.2. This Surveillance verifies the freedom of the CRAs to move, and may require a shutdown bank group to move below the insertion limits specified in the COLR, which would normally violate the LCO. This Note applies to each CRA shutdown bank group as the group is moved below the insertion limit to perform the Surveillance. This Note is not applicable should a malfunction stop performance of the Surveillance. Note that the CRA group alignment limits of LCO 3.1.4 remain applicable to the CRAs in the shutdown bank group being exercised while performing this Surveillance.

NuScale B 3.1.5-3 Draft Revision 3.0

Shutdown GroupBank Insertion Limits B 3.1.5 BASES ACTIONS A.1.1, A.1.2, and A.2 When one or more CRA shutdown bank groups CRA is not within insertion limits, 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months are allowed to restore the CRA shutdown bank groups CRA to within insertion limits. This is necessary because the available SDM may be significantly reduced with CRAone shutdownshowdown bank groups CRA not within their insertion limits.

Also, verification of the SDM or initiation of boration within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is required, since the SDM in MODE 1 is continuously monitored and adhered to, in part, by the CRA regulatingcontrol and shutdown groupbank insertion limits (see LCO 3.1.1).

The allowed Completion Time of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months provides an acceptable time for evaluating and repairing minor problems without allowing the unit to remain in an unacceptable condition for an extended period of time.

B.1 If the Required Actions and associated Completion Times are not met, the unit must be brought to a MODE where the LCO is not applicable.If the CRA shutdown bank groups cannot be restored to within their insertion limits within two hours, the unit must be brought to a MODE where the LCO is not applicable. The allowed Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable for reaching the required MODE from full power conditions in an orderly manner.

SURVEILLANCE SR 3.1.5.1 REQUIREMENTS Verification that the CRAs of each shutdown bank group is are within its insertion limits prior to an approach to criticality ensures that when the reactor is critical, or being taken critical, the shutdown bank groups will be available to shut down the reactor, and the required SDM will be maintained following a reactor trip. This SR and Frequency ensure that the CRA shutdown bank groups is are withdrawn before the CRA regulating bank groups are withdrawn during a unit startup.

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

REFERENCES 1. 10 CFR 50, Appendix A, GDC 10 and GDC 26.

2. 10 CFR 50.46.

3. FSAR, Chapter 15, Transient and Accident Analysis.

NuScale B 3.1.5-4 Draft Revision 3.0

Regulating GroupBank Insertion Limits B 3.1.6 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.6 Regulating BankGroup Insertion Limits BASES BACKGROUND The insertion limits of the regulating bank control rod assemblies (CRAs) are initial assumptions in the safety analyses that assume rod insertion upon reactor trip. The insertion limits directly affect core power and fuel burnup distributions, assumptions of available SDM, and initial reactivity insertion rate.

The applicable criteria for these reactivity and power distribution design requirements are 10 CFR 50, Appendix A, GDC 10, Reactor Design, GDC 26, Reactivity Control System Redundancy and Protection, GDC 28, Reactivity Limits (Ref. 1) and 10 CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors (Ref. 2). Limits on CRA regulating bank group CRA insertion have been established, and all regulating bank group CRA positions are monitored and controlled during power operation to ensure that the power distribution and reactivity limits defined by the design power peaking, ejected CRA worth, and SDM limits are preserved.

The 16 CRAs control rod assemblies (CRAs) are divided among the two regulating bank groups and twohe shutdown bank groups, with each group consisting of four CRAs in radially symmetric core locations. The regulating bank consists of two groups of four CRAs that are electrically paralleled to step simultaneously. See LCO 3.1.4, Rod Group Alignment Limits, for regulating and shutdown CRA rod OPERABILITY and alignment requirements, and LCO 3.1.7, Rod Position Indication, for CRA position indication requirements.

The regulating bank group insertion limits are specified in the COLR.

The Each CRA of a regulating bank groups are is required to be at or above the its regulating bank group insertion limits lines, as well as within its CRA group alignment limits.

The CRA regulating bank groups CRAs are used for precise reactivity control of the reactor. The positions of the CRAs in a regulating bank group are CRAs is normally controlled automatically by the Module Control System (MCS) together as a group of four CRAs;, but a regulating bank groups CRAs can also be manually controlled both individually and as a group. They CRA regulating bank groups are capable of changing core reactivity very quickly (compared to borating or diluting).

NuScale B 3.1.6-1 Draft Revision 3.0

Regulating GroupBank Insertion Limits B 3.1.6 BASES BACKGROUND (continued)

The power density at any point in the core must be limited so that the fuel design criteria are maintained. Together, LCO 3.1.4, Rod Group Alignment Limits, LCO 3.1.5, Shutdown GroupBank Insertion Limits, LCO 3.1.6, Regulating GroupBank Insertion Limits, LCO 3.2.1, Enthalpy Rise Hot Channel Factor (FH), and LCO 3.2.2, AXIAL OFFSET (AO) provide limits on control component operation and on monitored process variables which ensure that the core operates within the fuel design criteria.

The shutdown and regulating groupbank insertion and alignment limits and power distribution limits are process variables that together characterize and control the three dimensional power distribution of the reactor core. Additionally, the regulating groupbank insertion limits control the reactivity that could be added in the event of a rod ejection accident, and the shutdown and regulating groupbank insertion limits assure the required SDM is maintained.

Operation within the subject LCO limits will prevent fuel cladding failures that would breach the primary fission product barrier and release fission products to the reactor coolant in the event of a loss of coolant accident (LOCA), loss of flow, ejected CRA, or other accident requiring termination by a Reactor Trip System (RTS) trip function.

APPLICABLE The regulating groupbank insertion limits, FH, and AO LCOs are SAFETY required to prevent power distributions that could result in fuel cladding ANALYSES failures in the event of a LOCA, loss of flow, ejected CRA, or other accident requiring termination by an RTS trip function.

The acceptance criteria for addressing shutdown and regulating bank group insertion limits and inoperability or misalignment are that:

a. With the most reactive CRA stuck out there will be no violations of either:

1. specified acceptable fuel design limits; or

2. Reactor Coolant System (RCS) pressure boundary integrity; and

b. The core remains subcritical after design basis events with all CRAs fully inserted in the core.

NuScale B 3.1.6-2 Draft Revision 3.0

Regulating GroupBank Insertion Limits B 3.1.6 BASES APPLICABLE SAFETY ANALYSES (continued)

As such, the CRA shutdown and regulating groupbank insertion limits affect safety analysis involving core reactivity and power distributions (Ref. 3).

The SDM requirement is ensured by limiting the control and shutdown and regulating bankgroup insertion limits so that allowable inserted worth of the CRAs is such that sufficient reactivity is available in the CRAs to shut down the reactor to hot zero power with a reactivity margin which assumes the maximum worth CRA remains fully withdrawn upon trip (Ref. 3).

Operation at the insertion limits or AO limits may approach the maximum allowable linear heat generation rate or peaking factor.

Operation at the insertion limit may also indicate the maximum ejected CRA worth could be equal to the limiting value in fuel cycles that have sufficiently high ejected CRA worth.

The regulating and shutdown and regulating bankgroup insertion limits ensure that safety analyses assumptions for SDM, ejected rod worth, and power distribution peaking factors are preserved (Ref. 3).

The insertion limits satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii) in that they are initial conditions assumed in the safety analysis.

LCO The limits on regulating groupbank physical insertion as defined in the COLR, must be maintained because they serve the function of preserving power distribution, ensuring that the SDM is maintained, ensuring that ejected CRA worth is maintained, and ensuring adequate negative reactivity insertion is available on trip.

The LCO is modified by a Note indicating the LCO requirement is not applicable to control groups being inserted while performing SR 3.1.4.2. This SR verifies the freedom of the rods to move, and may require the control group to move below the LCO limits, which would normally violate the LCO. This Note applies to each control group as it is moved below the insertion limit to perform the SR. This Note is not applicable should a malfunction stop performance of the SR.

NuScale B 3.1.6-3 Draft Revision 3.0

Regulating GroupBank Insertion Limits B 3.1.6 BASES APPLICABILITY The regulating groupbank physical insertion limits shall be maintained with the reactor in MODE 1 when keff is 1.0. These limits must be maintained since they preserve the assumed power distribution, ejected CRA worth, SDM, and reactivity rate insertion rate assumptions. Applicability in MODE 1 with keff < 1.0, and MODES 2, 3, 4, and 5 is not required, since neither the power distribution nor ejected CRA worth assumptions would be exceeded in these MODES.

The applicability requirements have been modified by a Note indicating the LCO requirements are suspended during the performance of SR 3.1.4.2. This SR verifies the freedom of the rods to move, and requires the regulating group to move below the LCO limits, which would violate the LCO.The Applicability is modified by a Note indicating the LCO requirement is not applicable to CRA groups being inserted while performing SR 3.1.4.2. This SR verifies the freedom of the CRAs to move, and may require the regulating bank group to move below the LCO limits, which would normally violate the LCO.

This Note applies to each regulating bank group as it is moved below the insertion limit to perform the SR. This Note is not applicable should a malfunction stop performance of the SR.

ACTIONS A.1.1, A.1.2, and A.2 When the one or more regulating bank groups isare outside the acceptancenot within insertion limits, they must be restored to within those limits. This restoration can occur in two ways:

a. Reduce power to be consistent with rod CRA regulating bank group positions; or

b. Moving rods CRA regulating bank groups to be consistent with power.

Also, verification of SDM or initiation of boration to regain SDM is required within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , since the SDM in MODE 1 with keff is 1.0 is normally ensured by adhering to the control regulating and shutdown groupbank insertion limits (see LCO 3.1.1, "Shutdown Margin (SDM))

has been upset.

The allowed Completion Time of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for restoring the regulating bank groups to within insertion limits, provides an acceptable time for evaluating and repairing minor problems without allowing the unit to remain outside the insertion limitsin an unacceptable condition for an extended period of time.

NuScale B 3.1.6-4 Draft Revision 3.0

Regulating GroupBank Insertion Limits B 3.1.6 BASES ACTIONS (continued)

B.1 If the Required Actions cannot be completed within the associated Completion Times, the unit must be brought to a MODE where the LCO is not applicable.If the CRA regulating bank groups cannot be restored to within their insertion limits within two hours, the unit must be brought to a MODE where the LCO is not applicable. The allowed Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable for reaching the required MODE from full power conditions in an orderly manner.

SURVEILLANCE SR 3.1.6.1 REQUIREMENTS Verification of the regulating groupbank insertion limits is sufficient to detect regulating bank groups that may be approaching the insertion limits.

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

REFERENCES 1. 10 CFR 50, Appendix A, GDC 10, GDC 26, and GDC 28.

2. 10 CFR 50.46.

3. FSAR, Chapter 15, Transient and Accident Analyses.

NuScale B 3.1.6-5 Draft Revision 3.0

Rod Position Indication B 3.1.7 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.7 Rod Position Indication BASES BACKGROUND According to GDC 13 (Ref. 1), instrumentation to monitor variables and systems over their operating ranges during normal operation, anticipated operational occurrences (AOOs), and accident conditions must be OPERABLE. LCO 3.1.7 is required to ensure OPERABILITY of the control rod position indicators to determine control rod positions and thereby ensure compliance with the control rod alignment and power-dependent insertion limits (PDIL).

The OPERABILITY, including position indication, of the shutdown and regulating groupbank control rod assemblies (CRAs) is an initial assumption in the safety analyses that assume CRA insertion upon reactor trip. Maximum CRA misalignment is an initial assumption in the CRA misalignment safety analysis that directly affects core power distributions and assumptions of available shutdown margin (SDM).

CRA position indication is required to assess OPERABILITY and misalignment.

Mechanical or electrical failures may cause a CRA to become inoperable or to become misaligned from its group. CRA inoperability or misalignment may cause increased power peaking due to the asymmetric reactivity distribution and a reduction in the total available CRA worth for reactor shutdown. Therefore, CRA alignment and OPERABILITY are related to core operation in design power peaking limits and the core design requirement of a minimum SDM.

Limits on CRA alignment and OPERABILITY have been established, and CRA positions are monitored and controlled during power operation to aid compliance with the power distribution and reactivity limits defined by the design power peaking and SDM limits are preserved.

Sixteen CRAs are arranged in four symmetrical groups. Two shutdown bank groups of four CRAs each, and two regulating bank groups of four CRAs each.

CRAs are moved out of the core (up or withdrawn) or into the core (down or inserted) by their control rod drive mechanisms (CRDMs).

The CRAs are divided among the regulating bank groups and shutdown bank groups.

NuScale B 3.1.7-1 Draft Revision 3.0

Rod Position Indication B 3.1.7 BASES BACKGROUND (continued)

The axial position of shutdown bank CRAs and regulating bankgroup CRAs are determined by two separate and independent systemsmeans: the Counter Position Indicatorsion System (CPIs)

(commonly called bankgroup step counters) and the Rod Position Indicatorsion (RPIs) System.

The Counter Position IndicationCPI counts the commands sent to the CRDM gripper coils from the Control Rod Drive System (CRDS) that moves the CRAs. There is one step counter for each CRDM. The CRA Position Indication SystemCPI is considered highly precise

(+/- 1 step or +/- {3/8} inch). If a CRA does not move one step for each command signal, the step counter will still count the command and incorrectly reflect the position of the CRA.

The RPI function of the CRDS provides a highly accurate indication of actual CRA position, but at a lower precision than the step counters.

This system is based on inductive analog signals from a series of coils spaced along a hollow tube with a center to center distance of 1.125 inches, which is equivalent to 3 steps. To increase the reliability of the system, the inductive coils are alternately connected to two data systems. Thus, if one system fails, the RPI will go on half accuracy with an effective coil spacing of 2.25 inches, which is 6 steps.

Therefore, the normal indication accuracy of the RPIs System is +/- 3 steps (+/- 1.125 inches), and the accuracy with one channel of RPI out-of-service is +/- 6 steps

(+/- 2.25 inches).

APPLICABLE The regulating and shutdown bank groups CRA position accuracy is SAFETY essential during power operation. Power peaking, ejected CRA worth, ANALYSES or SDM limits may be violated in the event of a Design Basis Accident (Ref. 2), with regulating or shutdown bankgroup CRAs operating outside their limits undetected. Therefore, the acceptance criteria for CRA position indication is that CRA positions must be known with sufficient accuracy in order to verify the core is operating within the group sequence, overlap, design peaking limits, ejected CRA worth, and within minimum SDM (LCO 3.1.5, Shutdown BankGroup Insertion Limits, LCO 3.1.6, Regulating Bankgroup Insertion Limits). The CRA positions must also be known in order to verify the alignment limits are preserved (LCO 3.1.4, Rod Group Alignment Limits). CRA positions are continuously monitored to provide operators with information that assures the unit is operating within the bounds of the accident analysis assumptions.

NuScale B 3.1.7-2 Draft Revision 3.0

Rod Position Indication B 3.1.7 BASES APPLICABLE SAFETY ANALYSES (continued)

The CRA position indicator channels satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii). The control rod position indicators monitor CRA position, which is an initial condition of the accident.

LCO LCO 3.1.7 specifies that the RPIs System and the Counter Position Indication SystemCPI be OPERABLE for each CRA. For the CRA position indicators to be OPERABLE requires meeting the SR of the LCO and the following:

a. The RPI System indicates within 6 steps of the CRA counter demand position indicator as required by LCO 3.1.4, Rod Group Alignment Limits;

b. For the RPIs System there are no failed coils; and

c. The Counter Position Indication SystemCPI has been calibrated either in the fully inserted position or to the RPI System.

The 6 step agreement limit between the Rod Position Indication SystemRPIs and the CPI system indicates that the Rod Position Indication SystemRPI is adequately calibrated and can be used for indication of the measurement of CRA position.

A deviation of less than the allowable limit given in LCO 3.1.4 in position indication for a single CRA ensures high confidence that the position uncertainty of the corresponding CRA group is within the assumed values used in the analysis (that specified CRA groupbank insertion limits).

These requirements provide adequate assurance that CRA position indication during power operation and PHYSICS TESTS is accurate, and that design assumptions are not challenged.

OPERABILITY of the position indicator channels ensures that inoperable, misaligned, or mispositioned CRAs can be detected.

Therefore, power peaking, ejected CRA worth, and SDM can be controlled within acceptable limits.

APPLICABILITY The requirements on the RPI and step counters are only applicable in MODE 1 (consistent with LCOs 3.1.4, 3.1.5, and 3.1.6), because this is the only MODE in which power is generated, and the OPERABILITY and alignment of CRAs has the potential to affect the safety of the unit.

In the shutdown MODES, the OPERABILITY of the shutdown and NuScale B 3.1.7-3 Draft Revision 3.0

Rod Position Indication B 3.1.7 BASES APPLICABILITY (continued) regulating groupsbanks has the potential to affect the required SDM, but this effect can be compensated for by an increase in the boron concentration of the Reactor Coolant System (RCS).

ACTIONS The ACTIONS table is modified by a Note indicating that a separate Condition entry is allowed for each inoperable counter position indicatorCPI and each RPI indicator. This is acceptable because the Required Actions for each Condition provide appropriate compensatory actions for each inoperable position indicator.

A.1 When one RPI train per CRDM fails, the position of the CRA can still be determined by use of the In-Core Instrumentation System (ICIS).

Normal power operation does not require excessive movement of groups. If a group has been significantly moved, the Actions of B.1 or B.2 below are required. Therefore, verification of CRA position within the Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is adequate to allow continued full power operation, since the probability of simultaneously having a CRA significantly out of position and an event sensitive to that CRA position is small.

B.1, B.2, and B.3 When more than one RPI train per CRA fails, additional actions are necessary to ensure that acceptable power distribution limits are maintained, minimum SDM is maintained, and the potential effects of CRA misalignment on associated accident analyses are limited.

Placing the rod control function in manual mode ensures unplanned CRA motion will not occur. Together with the position determination available via the ICIS, this will minimize the potential for CRA misalignment. The immediate Completion Time for placing the Rod Control function in manual mode reflects the urgency with which unplanned rod motion must be prevented while in this Condition.

The position of the CRAs may be determined indirectly by use of the ICIS.

NuScale B 3.1.7-4 Draft Revision 3.0

PHYSICS TESTS Exceptions B 3.1.8 BASES BACKGROUND (continued)

The typical PHYSICS TESTS performed for reload fuel cycles (Ref. 4) in MODE 1 at < 5% RTP are listed below:

a. Critical Boron Concentration - Control Rods Withdrawn;

b. Control Rod Worth; and

c. Isothermal Temperature Coefficient (ITC).

These tests are initiated in MODE 1 at < 5% RTP. These and other supplementary tests may be required to calibrate the nuclear instrumentation or to diagnose operational problems. These tests may cause the operating controls and process variables to deviate from their LCO requirements during their performance.

a. The Critical Boron Concentration - Control Rods Withdrawn Test measures the critical boron concentration at hot zero power (HZP).

With rods out, the lead control group is at or near its fully withdrawn position. HZP is where the core is critical (keff = 1.0), and the Reactor Coolant System (RCS) is at design temperature and pressure for zero power. Performance of this test should not violate any of the referenced LCOs.

b. The Control Rod Worth Test is used to measure the reactivity worth of selected rod groups. This test is performed at HZP and has four alternative methods of performance. The first method, the Boron Exchange Method, varies the reactor coolant boron concentration and moves the selected regulating bank group in response to the changing boron concentration. The reactivity changes are measured with a reactivity computer. This sequence is repeated for the remaining regulating bank groups. The second method, the Rod Swap Method, measures the worth of a predetermined reference group using the Boron Exchange Method above. The reference group is then nearly fully inserted into the core. The selected group is then inserted into the core as the reference group is withdrawn. The HZP critical conditions are then determined with the selected group fully inserted into the core. The worth of the selected group is calculated based on the position of the reference group with respect to the selected group. This sequence is repeated as necessary for the remaining regulating groups. The third method, the Boron Endpoint Method, moves the selected regulating bank group over its entire length of travel while varying the reactor coolant boron concentration to maintain HZP criticality. The difference in boron concentration is the worth of the NuScale B 3.1.8-2 Draft Revision 3.0

PHYSICS TESTS Exceptions B 3.1.8 BASES BACKGROUND (continued) selected regulating bank group. This sequence is repeated for the remaining regulating groups. The fourth method, Dynamic Rod Worth Measurement (DRWM), moves each group, individually, into the core to determine its worth. The group is dynamically inserted into the core while data is acquired from the excore channel. While the group is being withdrawn, the data is analyzed to determine the worth of the group. This is repeated for each regulating bank and shutdown bank group. Performance of this test will violate LCO 3.1.4, Rod Group Alignment Limits, LCO 3.1.5, Shutdown BankGroup Insertion Limit, or LCO 3.1.6, Regulating BankGroup Insertion Limits.

c. The ITC Test measures the ITC of the reactor. This test is performed at HZP. The method is to vary the RCS temperature in a slow and continuous manner. The reactivity change is measured with a reactivity computer as a function of the temperature change.

The ITC is the slope of the reactivity versus the temperature plot.

The test is repeated by reversing the direction of the temperature change and the final ITC is the average of the two calculated ITCs.

Performance of this test should not violate any of the referenced LCOs.

APPLICABLE The fuel is protected by LCOs that preserve the initial conditions of the SAFETY core assumed during the safety analyses. The methods for ANALYSES development of the LCOs that are excepted by this LCO are described in the [NuScale Reload Safety Evaluation Methodology report]

(Ref. 5). The above mentioned PHYSICS TESTS, and other tests that may be required to calibrate nuclear instrumentation or to diagnose operational problems, may require the operating control or process variables to deviate from their LCO limitations.

FSAR Chapter 14 defines requirements for initial testing of the facility, including low power PHYSICS TESTS. FSAR Sections 14.2.10.3 and 14.2.10.4 (Ref. 6) summarize the initial criticality and low power tests.

Requirements for reload fuel cycle PHYSICS TESTS are defined in ANSI/ANS-19.6.1-2011 (Ref. 4). Although these PHYSICS TESTS are generally accomplished within the limits for the LCOs, conditions may occur when one or more LCOs must be suspended to make completion of PHYSICS TESTS possible or practical. This is acceptable as long as the fuel design criteria are not violated. When one or more of the requirements specified in:

LCO 3.1.3, Moderator Temperature Coefficient (MTC);

NuScale B 3.1.8-3 Draft Revision 3.0

PHYSICS TESTS Exceptions B 3.1.8 BASES APPLICABLE SAFETY ANALYSES (continued)

LCO 3.1.4, Rod Group Alignment Limits; LCO 3.1.5, Shutdown BankGroup Insertion Limit; and LCO 3.1.6, Regulating BankGroup Insertion Limits are suspended for PHYSICS TESTS, the fuel design criteria are preserved as long as the power level is limited to 5% RTP and SDM is within the limits provided in the COLR.

PHYSICS TESTS include measurement of core nuclear parameters or the exercise of control components that affect process variables. Also involved are the movable control components (regulating and shutdown CRArods), which are required to shut down the reactor. The limits for these variables are specified for each fuel cycle in the COLR.

As described in LCO 3.0.7, compliance with Test Exception LCOs is optional, and therefore no criteria of 10 CFR 50.36(c)(2)(ii) apply. Test Exception LCOs provide flexibility to perform certain operations by appropriately modifying requirements of other LCOs. A discussion of the criteria satisfied for the other LCOs is provided in their respective Bases.

LCO This LCO allows the reactor parameters of MTC to be outside their specified limits. In addition, it allows selected regulating and shutdown rods to be positioned outside of their specified alignment and insertion limits. Operation beyond specified limits is permitted for the purpose of performing PHYSICS TESTS and poses no threat to fuel integrity, provided the SRs are met.

The requirements of LCO 3.1.3, LCO 3.1.4, LCO 3.1.5, and LCO 3.1.6 may be suspended during the performance of PHYSICS TESTS provided:

a. SDM is within the limits provided in the COLR; and

b. THERMAL POWER is 5% RTP.

APPLICABILITY This LCO is applicable when performing low power PHYSICS TESTS.

The Applicability is stated as During PHYSICS TESTS initiated in MODE 1. Should the THERMAL POWER exceed 5% RTP, Required Action B.1 requires termination of critical operations by immediately opening the reactor trip breakers.

NuScale B 3.1.8-4 Draft Revision 3.0

PHYSICS TESTS Exceptions B 3.1.8 BASES ACTIONS A.1 and A.2 If the SDM requirement is not met, boration must be initiated promptly.

A Completion Time of 15 minutes is adequate for an operator to correctly align and start the required systems and components. The operator should begin boration with the best source available for the plant conditions. Boration will be continued until SDM is within limit.

Suspension of PHYSICS TESTS exceptions requires restoration of each of the applicable LCOs to within specification.

B.1 When THERMAL POWER is > 5% RTP, the only acceptable action is to open the reactor trip breakers (RTBs) to prevent operation of the reactor beyond its design limits. Immediately opening the RTBs will shut down the reactor and prevent operation of the reactor outside of its design limits.

SURVEILLANCE SR 3.1.8.1 REQUIREMENTS Verification that the THERMAL POWER is 5% RTP will ensure that the unit is not operating in a condition that could invalidate the safety analyses.

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

SR 3.1.8.2 The SDM is verified by performing a reactivity balance calculation, considering the following reactivity effects:

a. RCS boron concentration;

b. Regulating bank group positions;

c. RCS average temperature;

d. Fuel burnup based on gross thermal energy generation;

e. Xenon concentration;

f. Samarium concentration; and

g. Isothermal temperature coefficient (ITC).

NuScale B 3.1.8-5 Draft Revision 3.0

FH B 3.2.1 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.1 Enthalpy Rise Hot Channel Factor (FH)

BASES BACKGROUND The purpose of this LCO is to establish limits on the power density at any point in the core so that the fuel design criteria are not exceeded and the accident analysis assumptions remain valid. Control of the core power distribution with respect to these limits ensures that local conditions in the fuel rods and coolant channels do not challenge core integrity at any location during either normal operation or a postulated accident analyzed in the safety analyses.

FH is defined as the ratio of the maximum integrated rod power within the core to the average rod power. Therefore, FH is a measure of the maximum total power produced in a fuel rod.

FH is sensitive to fuel loading patterns, regulating bank group insertion, and fuel burnup. FH typically increases with regulating bank group insertion and typically decreases with fuel burnup.

FH is not directly measurable but is inferred from a power distribution map obtained with the fixed incore detector system. Specifically, the measurements taken from the fixed incore instrument system are analyzed by a computer to determine FH. This value is calculated continuously with operator notification on unexpected results and validated by engineering in accordance with the surveillance frequency.

The COLR provides peaking limits that ensure that the safety analysis values for critical heat flux (CHF) are not exceeded for normal operation, operational transients, and any transient condition arising from analyzed events. The safety analysis precludes CHF and is met by limiting the minimum critical heat flux ratio (MCHFR) to that value defined in the COLR. All transient events are assumed to begin with an FH value that satisfies the LCO requirements.

Operation outside the LCO limits may produce unacceptable consequences if an event occurs. The CHF safety analysis ensures that there is no overheating of the fuel that results in possible cladding perforation with the release of fission products to the reactor coolant.

NuScale B 3.2.1-1 Draft Revision 3.0

FH B 3.2.1 BASES APPLICABLE Limits on FH preclude core power distributions that exceed fuel design SAFETY limits.

ANALYSES There must be at least 95% probability at the 95% confidence level (the 95/95 CHF criterion) that the hottest fuel rod in the core does not experience a CHF condition.

The limits on FH ensure that the safety analysis values for CHF are not exceeded for normal operation, operational transients, and any transient condition arising from analyzed events. The safety analysis precludes CHF and is met by limiting the MCHFR to that value defined in the COLR.

This value provides a high degree of assurance that the hottest fuel rod in the core does not experience a CHF condition.

The allowable FH limit increases with decreasing power level. This functionality in FH is included in the analyses that provide the Reactor Core Safety Limits (SLs) of SL 2.1.1. Therefore, any CHF events in which the calculation of the core limits is modeled implicitly use this variable value of FH in the analyses. Likewise, all transients that may be CHF limited are assumed to begin with an initial FH as a function of power level defined by the COLR limit equation.

The fuel is protected in part by Technical Specifications, which ensure that the initial conditions assumed in the safety and accident analyses remain valid.

FH is measured periodically using the fixed incore detector system.

Measurements are generally taken with the core at, or near, steady state conditions. Core monitoring and control under transient conditions are accomplished by operating the core within the limits of the LCOs on AO and GroupBank Insertion Limits.

FH satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO FH shall be maintained within the limits of the relationship provided in the COLR.

The FH limit identifies the coolant flow channel with the maximum enthalpy rise. This channel has the least heat removal capability and thus the highest probability for a CHF condition.

The limiting value of FH, described by the equation contained in the COLR, is the design radial peaking limit used in the safety analyses.

NuScale B 3.2.1-2 Draft Revision 3.0

AO B 3.2.2 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.2 AXIAL OFFSET (AO)

BASES BACKGROUND The purpose of this LCO is to establish limits on the values of AO in order to limit the amount of axial power distribution skewing to either the top or bottom of the core. By limiting the amount of power distribution skewing, core peaking factors are consistent with the assumptions used in the safety analyses. Limiting power distribution skewing over time also minimizes the xenon distribution skewing, which is a significant factor in axial power distribution control.

The AO limits are selected by considering a range of axial xenon distributions that may occur as a result of large variations of the AO.

Subsequently, power peaking factors and power distributions are examined to ensure that the postulated event limits are met. Violation of the AO limits invalidate the conclusions of the accident and transient analyses with regard to fuel cladding integrity. (Ref. 1)

The in-core instrumentation system's neutron detectors are arranged equally spaced radially and axially throughout the core. This neutron detector arrangement promotes an accurate indication for the module control system to analyze core power distributions and will be used to monitor AO.

APPLICABLE The AO is a measure of the axial power distribution skewing to either the SAFETY top or bottom half of the core. The AO is sensitive to many core related ANALYSES parameters such as regulating bank group positions, core power level, axial burnup, axial xenon distribution, reactor coolant temperature, and boron concentration.

The allowed range of the AO is used in the nuclear design process to confirm that operation within these limits produces core peaking factors and axial power distributions that meet safety analysis requirements.

The limits on the AO ensure that the bounding axial power distribution is not exceeded during either normal operation or in the event of xenon redistribution following power changes. The limits on the AO also restrict the range of power distributions that are used as initial conditions in the analyses of anticipated operational occurrences (AOO), infrequent events (IE), and accidents. This ensures that the fuel cladding integrity is maintained for these postulated accidents. The most important AOO is the Control Rod Misoperation - Single Rod Withdrawal. The most NuScale B 3.2.2-1 Draft Revision 3.0

RCS Pressure, Temperature, and Flow Resistance CHF Limits B 3.4.1 BASES APPLICABLE SAFETY ANALYSES (continued)

The NSP2 and NSP4 correlation limits are used for comparison to conditions representative of normal operation, operational transients, anticipated operational occurrences, and accidents other than events that are initiated by rapid reductions in primary system inventory. The Hench-Levy correlation is used to evaluate events for which analyses postulate a rapid reduction in primary system inventory. An assumption for the analysis of these events is that the core power distribution is within the limits of LCO 3.1.6, Regulating GroupBank Insertion Limits; LCO 3.2.1, "Enthalpy Rise Hot Channel Factor (FH)," and LCO 3.2.2, AXIAL OFFSET (AO).

The flow resistance in the RCS directly affects the reactor coolant natural circulation flow rate established by THERMAL POWER, RCS pressure, and RCS temperature. The safety analyses assume flow rates that are based on a conservative value of flow resistance through the RCS.

Therefore the resistance must be verified to ensure that the assumptions in the safety analyses remain valid.

The pressurizer pressure operating limit and the RCS averagecold temperature limit specified in the COLR, as shown on the Analytical Design Operating LimitsThermal Margins Limit Map, correspond to operatinganalytical limits, with an allowance for steady state fluctuations and measurement errors. These are the analytical initial conditions assumed in transient and LOCA analyses.

The RCS CHF parameters satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO This LCO specifies limits on the monitored process variables, pressurizer pressure and RCS cold temperature to ensure the core operates within the limits assumed in the safety analyses. It also specifies the limit on RCS flow resistance to ensure that the RCS flow is consistent with the flow assumed in the safety analyses. These variables are contained in the COLR to provide operating and analysis flexibility from cycle to cycle.

Operating within these limits will result in meeting CHFR criterion in the event of a CHF-limited transient.

NuScale B 3.4.1-2 Draft Revision 3.0

Response to Request for Additional Information Docket No.52-048 eRAI No.: 9634 Date of RAI Issue: 11/29/2018 NRC Question No.: 16-60-22

21. In Revision 2 of DCA part 4, GTS Bases page B 3.1.6-1, Background section:

21.1 The applicant is requested to clarify the second paragraph, second sentence as indicated:

... Limits on CRA regulating group CRA insertion have been established, and all regulating group CRA positions are monitored and controlled during power operation to ensure that the power distribution and reactivity limits defined by the design power peaking, ejected CRA worth, and SDM limits are preserved.

21.2 The applicant is requested to clarify the third, fourth, and fifth paragraphs, as indicated:

3rd The control rod assemblies (CRAs) 16 CRAs are divided among the two regulating groups and the two shutdown groups, with each group consisting of four CRAs in radially symmetric core locations. The regulating bank consists of two groups of four CRAs, which that are electrically paralleled to step simultaneously. See LCO 3.1.4, "Rod Group Alignment Limits,"

for regulating and shutdown rod CRA OPERABILITY and alignment requirements, and LCO 3.1.7, "Rod Position Indication," for CRA position indication requirements.

4th The regulating group insertion limits are specified in the COLR. Each CRA of a The regulating groups are is required to be at or above the its regulating group insertion limit lines, as well as within its CRA group alignment limits.

5th The CRA regulating group CRAs are used for precise reactivity control of the reactor.

The positions of the CRAs in a regulating group are group CRAs is normally controlled automatically by the Module Control System (MCS) together as a group of four CRAs;, but a NuScale Nonproprietary

regulating group's CRAs can also be manually controlled, both individually and as a group. They The CRA regulating groups are capable of changing core reactivity very quickly (compared to borating or diluting).

NuScale Response:

The Bases have been modified.