ML18200A443

LLC Submittal of Changes to Final Safety Analysis Report Section 3.9.4.4, Control Rod Drive System Operability Assurance Program and Section 1.8, Interfaces with Certified Design

ML18200A443
Person / Time
Site:	NuScale
Issue date:	07/19/2018
From:	Wike J NuScale
To:	Document Control Desk, Office of New Reactors
References
LO-0718-60969
Download: ML18200A443 (6)

Text

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NUSCALE

.:. POWER' LO-0718-60969 July 19, 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 Submittal of Changes to Final Safety Analysis Report Section 3.9.4.4, "Control Rod Drive System Operability Assurance Program" and Section 1.8, "Interfaces with Certified Design"

REFERENCES:

Letter from NuScale Power, LLC to Nuclear Regulatory Commission, "NuScale Power, LLC Submittal of the NuScale Standard Plant Design Certification Application , Revision 1," dated March 15, 2018 (ML18086A090)

During an April 23, 2018 closed teleconference with Omid Tabatabai and Nicholas Hansing of the NRC staff, NuScale Power, LLC (NuScale) discussed potential updates to Final Safety Analysis Report (FSAR) Section 3.9.4.4, "Control Rod Drive Operability Assurance Program."As a result of this discussion, NuScale changed FSAR Section 3.9.4.4 to restore information that had been inadvertently removed and made conforming changes to FSAR Section 1.8," lnterfaces with Certified Design." The Enclosure to this letter provides a mark-up of the FSAR pages incorporating revisions to FSAR Section 3.9.4.4 in redline/strikeout format. NuScale will include this change as part of Revision 2 to the NuScale Design Certification Application.

This letter makes no regulatory commitments or revisions to any existing regulatory commitments.

Please feel free to contact Marty Bryan at 541-452-7172 or at mbryan@nuscalepower.com if you have any questions.

Sincerely,

!}ft Jee Jennie wreu Manager, Licensing NuScale Power, LLC Distribution: Samuel Lee, NRC, OWFN-8G9A Gregory Cranston, NRC, OWFN-8G9A Marieliz Vera, NRC, OWFN-8G9A

Enclosure:

"Changes to NuScale Final Safety Analysis Report Section 3.9.4.4 , 'Control Rod Drive Operability Assurance Program' and Section 1.8, 'Interfaces with Certified Design "'

NuScale Power, LLC 1100 NE Circle Blvd , Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com

LO-0718-60969

Enclosure:

Changes to NuScale Final Safety Analysis Report Sections 3.9.4.4, Control Rod Drive Operability Assurance Program and 1.8, Interfaces with Certified Design NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com

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 3.9-1: A COL applicant that references the NuScale Power Plant design certification will provide the 3.9 applicable test procedures before the start of testing and will submit the test and inspection results from the comprehensive vibration assessment program for the NuScale Power Module, in accordance with Regulatory Guide 1.20.

COL Item 3.9-2: A COL applicant that references the NuScale Power Plant design certification will develop 3.9 design specifications and design reports in accordance with the requirements outlined under American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section III (Reference 3.9-1). A COL applicant will address any known issues through the reactor vessel internals reliability programs (i.e. Comprehensive Vibration Assessment Program, steam generator programs, etc.) in regards to known aging degradation mechanisms such as those addressed in Section 4.5.2.1.

COL Item 3.9-3: A COL applicant that references the NuScale Power Plant design certification will provide a 3.9 summary of reactor core support structure ASME service level stresses, deformation, and cumulative usage factor values for each component and each operating condition in conformance with ASME Boiler and Pressure Vessel Code Section III Subsection NG.

COL Item 3.9-4: A COL applicant that references the NuScale Power Plant design certification will submit a 3.9 Preservice Testing program for valves as required by 10 CFR 50.55a.

COL Item 3.9-5: A COL applicant that references the NuScale Power Plant design certification will establish an 3.9 Inservice Testing program in accordance with ASME OM Code and 10 CFR 50.55a.

COL Item 3.9-6: A COL applicant that references the NuScale Power Plant design certification will identify any 3.9 site-specific valves, implementation milestones, and the applicable ASME OM Code (and ASME OM Code Cases) for the preservice and inservice testing programs. These programs are to be consistent with the requirements in the latest edition and addenda of the OM Code incorporated by reference in 10 CFR 50.55a in accordance with the time period specified in 10 CFR 50.55a before the scheduled initial fuel load (or the optional ASME Code Cases listed in Regulatory Guide 1.192 incorporated by reference in 10 CFR 50.55a).

COL Item 3.9-7: Not Used.

COL Item 3.9-8: A COL applicant that references the NuScale Power Plant design certification will develop 3.9 specific test procedures to allow detection and monitoring of power-operated valve assembly performance sufficient to satisfy periodic verification design basis capability requirements.

COL Item 3.9-9: A COL applicant that references the NuScale Power Plant design certification will develop 3.9 specific test procedures to allow detection and monitoring of emergency core cooling system valve assembly performance sufficient to satisfy periodic verification of design basis capability requirements.

COL Item 3.9-10: A COL applicant that references the NuScale Power Plant design certification will verify that 3.9 evaluations are performed during the detailed design of the main steam lines, using acoustic resonance screening criteria and additional calculations as necessary (e.g., Strouhal number) to determine if there is a concern. The methodology contained in NuScale Comprehensive Vibration Assessment Program Technical Report, TR-0716-50439 is acceptable for this purpose.

The COL applicant will update Section 3.9.2.1.1.3 to describe the results of this evaluation.

COL Item 3.9-11: A COL applicant that references the NuScale Power Plant design certification will implement a 3.9 CRDS Operability Assurance Program that meets the requirements described in NUREG-0800, SRP 3.9.4, Revision 3, Acceptance Criteria II.4.

COL Item 3.10-1: A COL applicant that references the NuScale Power Plant design certification will develop and 3.10 maintain a site-specific seismic and dynamic qualification program.

COL Item 3.10-2: A COL applicant that references the NuScale Power Plant design certification will develop the 3.10 equipment qualification database and ensure equipment qualification record files are created for the structures, systems, and components that require seismic qualification.

COL Item 3.10-3: A COL applicant that references the NuScale Power Plant design certification will submit an 3.10 implementation program for Nuclear Regulatory Commission approval prior to the installation of the equipment that requires seismic qualification.

Tier 2 1.8-7 Draft Revision 2

NuScale Final Safety Analysis Report Mechanical Systems and Components Control Rod Reconnect After the reactor has been refueled and the plant is restored to the state it was in at the completion of the remote disconnect mechanism disengagement process, the remote engagement process (i.e., reconnect) begins. The control rod drive shaft reconnects to the CRA hub by lowering the outer shaft to retract the plug on the bottom of the center disconnect rod. Utilizing the stepping process, the control rod drive shaft is lowered into the CRA hub with the plug on the center disconnect rod extracted. The control rod drive shaft is then lowered an extra step to make sure the fingers on the coupling are fully inserted into the CRA hub. The next step is to insert the plug on the center disconnect rod to expand the fingers and lock them in place. To ensure the plug on the center disconnect rod completely inserts, the remote disconnect gripper releases the center disconnect rod and lets it fall, along with the spring assist.

RAI 03.09.04-1S1, RAI 03.09.04-2S1, RAI 03.09.04-4S1 To initiate the reconnect sequence, the control rod shaft is raised above the post-scram position. This is the same position from which the disconnect sequence was started. The disconnect gripper is activated to make the center disconnect rod stationary to remove the plug on the center disconnect rod from the locked 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 inadvertentintentional 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 Tier 2 3.9-41 Draft Revision 2

NuScale Final Safety Analysis Report Mechanical Systems and Components

• design pressure (RCS) - 2,100 psia

• normal operating pressure (RCS) - 1,850 psia

• design temperature (RCS) - 650 degrees Fahrenheit

• normal operating temperature (RCS) - 625 degrees Fahrenheit The CRDMs are designed for the loading combinations and loading values specified in Section 3.9.3.

The worth of the 16 CRA in conjunction with the CRDS trip function is sufficient to overcome a stuck rod event. In addition, design requirements have been established for clearances during seismic, thermal expansion and dynamic events.

3.9.4.4 Control Rod Drive System Operability Assurance Program The ability of the CRDS pressure housing components within the CRDMs to perform throughout the operating design life of 60 years is confirmed by the design report required by the ASME BPVC,Section III (Reference 3.9-1).

Although the NuScale CRDS is similar in design to the CRDSs of the currently operating fleet of PWRs, it has some unique features that include a longer control rod drive shaft (due to the presence of an integral SG and a pressurizer volume between the top of the core and the top of the RPV), and a remote disconnect mechanism. A prototype testing program was created that integrates the CRDM, the control rod drive shaft, the CRA, and the fuel assembly to demonstrate the acceptable mechanical functioning of a prototype CRDS. Rod drops under various conditions are tested and measured.

The testing of the prototype includes CRA drop time and misalignment testing and wear susceptibility assessment as described by Section 4.2.4.2.3.

The NuScale CRDS design is subject to an Operability Assurance Program. The CRDS Operability Assurance Program testing is composed of a series of tests designed to qualify the life cycle performance and endurance, including wear characteristics, of the CRDS in an operational environment. These tests include the number of position adjustments, the number of scrams and the expected thermal cycles. These tests also evaluate CRDS performance due to abnormal and accident events including an evaluation of the long control rod drive shaft during a seismic event. The minimum testing requirements and acceptance criteria for these tests utilize the configuration of the final design of the CRDM. This series of tests complies with the CRDS Operability Assurance Program requirements described in NUREG 0800, SRP 3.9.4, Revision 3, Acceptance Criteria II.4.

COL Item 3.9-11: A COL applicant that references the NuScale Power Plant design certification will implement a CRDS Operability Assurance Program that meets the requirements described in NUREG-0800, SRP 3.9.4, Revision 3, Acceptance Criteria II.4.

A series of production tests are performed on each CRDM that verifies the integrity of the pressure housing and the function of the CRDM. These tests include a hydrostatic test in accordance with the ASME BPVC Code,Section III, Division I, Subsection NB.

Tier 2 3.9-45 Draft Revision 2

NuScale Final Safety Analysis Report Mechanical Systems and Components The as-built CRDMs are subject to pre-operational testing that verifyies the sequencing of the operating coils and verifyies the design requirements are met for insertion, withdrawal, and drop times. A description of the initial startup test program is provided in Section 14.2.

In accordance with the technical specifications, the CRDMs are subjected periodically to partial-movement checks to demonstrate the operation of the CRDM and acceptable core power distribution. In addition, drop tests of the CRA are performed at each refueling as specified in Technical Specification Surveillance Requirement 3.1.4.3 to verify the ability to meet trip time requirements.

3.9.5 Reactor Vessel Internals The RVI assembly is comprised of several sub-assemblies which are located inside the RPV.

The RVI support and align the reactor core system, which includes the control rod assemblies (CRAs), support and align the control rod drive rods, and include the guide tubes that support and house the in-core instrumentation (ICI). In addition to performing these support and alignment functions, the RVI channels the reactor coolant from the reactor core to the steam generator (SG) and back to the reactor core.

The RVI primary functions are to:

• provide structures to support, properly orient, position, and seat the fuel assemblies to maintain the fuel in an analyzed geometry to ensure core cooling capability and physics parameters are met under all modes of operational and accident conditions

• provide support and properly align the CRDS without precluding full insertion of control rods under all modes of operational and accident conditions

• provide the flow envelope to promote natural circulation of the RCS fluid with consideration given to minimizing pressure losses and bypass leakage associated with the RVI, and to the flow of coolant to the core during refueling operations The RVI assembly is comprised of the following sub assemblies/items:

• core support assembly (CSA)

• lower riser assembly

• upper riser assembly

• flow diverter

• PZR spray nozzles The design and construction of both the core support structures and the internal structures that comprise the RVI comply with the requirements of ASME BPVC Section III, Division 1, Subsection NG. Safety-related structures and components are constructed and tested to quality standards commensurate with the importance of the safety-related functions to be performed, and designed with appropriate margins to withstand effects of normal operation, AOOs, natural phenomena such as earthquakes, and postulated accidents including LOCA, as discussed in GDC 1, 2, 4 and 10 and 10 CFR 50.55a.

Tier 2 3.9-46 Draft Revision 2