ML18348A917

LLC Supplemental Response to NRC Request for Additional Information No. 340 (Erai No. 9358) on the NuScale Design Certification Application

ML18348A917
Person / Time
Site:	NuScale
Issue date:	12/13/2018
From:	Rad Z NuScale
To:	Document Control Desk, Office of New Reactors
References
RAIO-1218-63846
Download: ML18348A917 (10)

Text

RAIO-1218-63846 December 13, 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 Supplemental Response to NRC Request for Additional Information No. 340 (eRAI No. 9358) on the NuScale Design Certification Application

REFERENCES:

1. U.S. Nuclear Regulatory Commission, "Request for Additional Information No. 340 (eRAI No. 9358)," dated January 26, 2018

2. NuScale Power, LLC Response to NRC "Request for Additional Information No. 340 (eRAI No.9358)," dated March 27, 2018

3. NuScale Power, LLC Supplemental Response to NRC "Request for Additional Information No. 340 (eRAI No. 9358)," dated September 13, 2018

4. NuScale Power, LLC Supplemental Response to NRC "Request for Additional Information No. 340 (eRAI No. 9358)," dated November 11, 2018 The purpose of this letter is to provide the NuScale Power, LLC (NuScale) supplemental response to the referenced NRC Request for Additional Information (RAI).

The Enclosure to this letter contains NuScale's supplemental response to the following RAI Question from NRC eRAI No. 9358:

• 03.06.02-17 This submittal provides a third supplemental response for RAI 9358 Question 03.06.02-17. A brief recap of this RAI question supplemental response history is included as introduction to the NuScale response.

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

If you have any questions on this response, please contact Marty Bryan at 541-452-7172 or at mbryan@nuscalepower.com.

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

RAIO-1218-63846 Distribution: Gregory Cranston, NRC, OWFN-8G9A Samuel Lee, NRC, OWFN-8G9A Marieliz Vera, NRC, OWFN-8G9A : NuScale Supplemental Response to NRC Request for Additional Information eRAI No. 9358 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com

RAIO-1218-63846 :

NuScale Supplemental Response to NRC Request for Additional Information eRAI No. 9358 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.: 9358 Date of RAI Issue: 01/26/2018 NRC Question No.: 03.06.02-17 In response to RAI 9187, Question 03.06.02-16, NuScale stated that the configuration of the RVVs and RRVs had changed from a welded connection to a bolted connection.

In that response, NuScale also referred to its response to RAI 8776, Question 15.06.06-5, to support NuScales position that high energy line breaks do not need to be postulated at the RVV and RRV connections to the RPV. Specifically, NuScale referred to Section III of the ASME BPV Code which defines piping system as an assembly of piping, piping supports, components, and, if applicable, components supports. Further, NuScale stated that while a piping system may include non-piping components such as a valve, a piping system must at least include piping. Moreover, NuScale stated that in the NuScale design, there is no piping between the Reactor Pressure Vessel (RPV) nozzles and Reactor Vent Valves (RVVs)/Reactor Recirculation Valves (RRVs), but rather only two non-piping components welded together. Therefore, NuScales position is that high energy line breaks do not need to be postulated at the RVV and RRV connections to the RPV.

The NRC staff disagreed with the above NuScales interpretation of the piping system as defined in the ASME Code. The NRC staffs interpretation is that a piping system is a system that includes any of the following, piping, piping supports, components, or components supports. This NRC staffs interpretation is consistent with the definition and scope of vessel and pipe as described by the ASME Companion Guide. As described in RAI 9187, Question 03.06.02-16, Companion Guide to the ASME Boiler and Pressure Vessel Code states that Paragraph U-1(a)(2) of ASME Section VIII-1 scope addresses pressure vessels that are defined as containers for the containment of pressure, internal or external and if the primary function of the pressure container is to transfer fluid from one point in the system to another, then the component should be considered as piping. Further, Paragraph 21.3.1.2 of the NuScale Nonproprietary

Companion Guide states that the vessel boundary ends at the face of the flange for bolted connections to piping, other pressure vessels, and mechanical equipment.

Accordingly, the NRC staff considers the boundary of the vessel to be at the [bolted flange connections between the RVV and RRV and the vessel]. Therefore, the staffs position is that RVV and RRV should be considered as part of the piping system and is the extremity of the affected piping system. As stated in BTP 3-4 Section 2A(iii) that breaks should be postulated at the terminal end of each piping run. Bolting the RVVs and RRVs to a flanged connect to the reactor vessel would be a terminal end connection.

For the NuScale RVV and RRV design, the NRC staffs key concern is that this bolted flange connection to the reactor vessel must not fail catastrophically, causing a loss-of-coolant accident. Operating experience from current reactors demonstrates that degradation and failure do occur at bolted connections in nuclear power plants. Electric Power Research Institute (EPRI) NP-5769, Degradation and Failure of Bolting in Nuclear Power Plants, dated April 1988, discusses various causes of bolting degradations and failures. The contributing factors to these incidents include stress corrosion cracking, boric acid corrosion, flow-induced vibration, improper torque/preload, and steam cutting. NUREG-1339, resolution of Generic Safety Issue 29: Bolting Degradation or Failure in Nuclear Power Plants, dated June 1990, discusses resolution of issues from this EPRI study. Specifically, it discusses NRCs evaluation of and exceptions to EPRI NP-5769. Further, Generic Letter (GL) 91-17, Bolting Degradation or Failure in Nuclear Power Plants, provides information on the resolution of GSI 29.

Per the response to RAI No. 8785, Question 15.06.05-1 and based on our previous interactions with NuScale, the staff understands that NuScale is not assuming a break at this location. There is precedent for not postulating breaks in certain locations where additional design and operational criteria provide assurance that this approach is acceptable. GDC 4 explicitly allows exclusion of certain pipe ruptures when the probability of fluid system piping rupture is extremely low- the basis used for leak-before-break as described in SRP Section 3.6.3, Leak-Before-Break Procedures. The specific guidelines included in SRP 3.6.3, are a deterministic fracture-mechanics-based approach. They are applicable for pipes only and cannot be directly applied to a bolted flange connection. However, the concept of demonstrating that leakage will be detected in time to ensure that the probability of gross failure is extremely low should be the same.

In addition, Section 2A(ii) of BTP 3-4 states that breaks need not be postulated in those portions of piping from containment wall to and including the inboard or outboard isolation valves (the break exclusion zone), provided they meet certain specific design criteria for stress and fatigue NuScale Nonproprietary

limits, welding, pipe length, guard pipe assemblies, and full volumetric examination of welds.

These existing break exclusion guidelines are for fluid system piping in the containment penetration area of current generation large light-water reactors and, therefore, are not directly applicable to NuScale.

If NuScale desires to treat the bolted connection of the RRVs and RVVs to a flange connected to the reactor vessel as a break exclusion area, then a justification for why this connection provides confidence that the probability of gross rupture is extremely low, must be provided for NRC staff review and acceptance. The justification will need to contain a discussion of the considerations outlined below.

1. Quantitative assessment of the probability of gross failure for the bolted flange connection

2. Specific design stress and fatigue limits

3. A comprehensive bolting integrity program in accordance with the recommendations and guidelines in NUREG-1339 (with additional detail provided in EPRI NP-5769, as referenced in NUREG-1339), as well as related NRC bulletins and generic letters

4. Local leakage detection (potentially similar in concept to leakage detection from reactor vessel heads) that will provide indication of leakage before gross bolt failure, such that the plant can shut down

5. Augmented inspection program requirements, which could include augmented procurement requirements for the bolting, ultrasonic in-service testing of the bolts of the bolted flange connection at some specific inspection frequency, periodic bolt replacement, etc.

The staff requests the applicant to clarify how they intend to treat the bolted connection as a break exclusion location and if so, provide justification with a discussion of the above considerations.

NuScale Nonproprietary

NuScale Response:

The following information supplements that provided in the earlier responses to RAI 9358 Question 03.06.02-17 as transmitted by NuScale letters RAIO-0318-59309, 3/27/2018 (initial response), RAIO-0918-61767, 9/13/2018 (Supplement 1), and RAIO-1118-62971, 11/15/2018 (Supplement 2).

A brief recap of the RAI 9358 Question 03.06.02-17 supplemental response history follows:

Supplement 1 - During a May 1, 2018 public call the NRC stated that the ECCS valve flange bolts should be designed to more conservative stress and fatigue criteria than the ASME code, similar to the break postulation criteria of the BTP 3-4. NuScale agreed to supplement its response to address the application of more conservative design criteria.

Also during the call NRC expressed concern with NuScale not performing ultrasonic testing (UT) examination of removed flange bolts, because of industry experience with VT1 inspection missing bolting flaws. NuScale subsequently agreed to supplement its RAI response either with additional justification for the VT1 versus UT bolt examinations or to change the VT1 exam to a UT exam.

And, the NRC questioned whether potential leakage from the bolted flange connection could disrupt ECCS system operations, including if induced vibrations could cause damage. NuScale was asked if bolted flange leakage is categorized and controlled by the Technical Specifications. NuScale agreed to further supplement its RAI response to address these additional concerns.

Supplement 2 - During a public clarification call on 10/16/2018, the NRC indicated that ECCS valve actuation should be considered normal operation for the ECCS valves, and the associated dynamic load evaluated to service level A or service level B. NuScale agreed to evaluate the NRC position and provide more information in a supplemental response.

Supplement 3 - During an October 31, 2018 public call the NRC noted that the supplemental response to RAI 9358 Question 03.06.02-17 (Supplement 1 above) indicated that RVVs and RRVs are within the scope of the NuScale CVAP. NuScale stated that RRVs were evaluated for susceptibility to acoustic resonance, and subsequently found to be acceptable. Therefore, due to their inclusion in the CVAP, ECCS valves do not require additional margin in CUF limits to account for possible vibration loading. Given this conclusion, NRC requested that a NuScale Nonproprietary

discussion be added to the CVAP technical report relative to this ECCS valve evaluation to acoustic resonance.

Response to Question 03.06.02-17 Supplement 3 -

During an accident, the Emergency Core Cooling System (ECCS) may be actuated, allowing flow through the ECCS valves. This flow is assessed according to the methodology of the CVAP for susceptibility to the six flow-induced vibration (FIV) mechanisms. The ECCS valves are an angle globe valve design, wherein flow makes a 90 degree turn and passes around the valve disc when the valve is open. The valve disc is not in direct cross flow and downstream structures (the valve 90 degree turn) are present to disrupt any potential vortices generated by the valve internals. The valve body is designed for reaction loads of valve discharge and seismic loads, and is therefore thick-walled relative to schedule 160 piping. It is not a bounding component for turbulent buffeting analysis. Due to the geometry in the ECCS valves, no FIV mechanisms are credible for through-valve flow.

• Fluid Elastic Instability: Not susceptible due to geometry (not an array of cylinders)

• Vortex Shedding: Not susceptible due to geometry (downstream structures are present to disrupt vortices)

• Turbulent Buffeting: Not a bounding structure for analysis (thick-walled)

• Acoustic Resonance: Not susceptible due to geometry (no flow-occluded cavities)

• Leakage Flow: Not susceptible due to geometry (no flexible structures)

• Galloping and Flutter: Not susceptible due to geometry (no non-circular cross sections with susceptible aspect ratios)

Impact on DCA:

The CVAP Technical Report TR-0716-50439 Section 2.3.6 has been revised as described in the response above and as shown in the markup provided in this response.

NuScale Nonproprietary

NuScale Comprehensive Vibration Assessment Program Technical Report TR-0716-50439-NP Draft Rev. 12 Figure 2-24 Flow diverter 2.3.6.4 Thermowells Within the primary coolant flow path, RCS temperature instruments are installed in thermowells located at the entrance to the SG tubes and in the downcomer.

Thermowells are welded to the RPV. Thermowells are also used to measure temperature in secondary side piping. In locations where they are used, thermowells extend into the flow path and are exposed to turbulent cross-flow conditions. Therefore, they are susceptible to VS and TB. Other FIV mechanisms are not applicable to these components.

2.3.6.5 Component and Instrument Ports Acoustic resonances due to the generation of vortices at closed branch lines are evaluated. Penetrations that create a hollow cavity and that are located in regions with adjacent flow are susceptible to AR. Cavities form acoustic standing waves which may be excited due to the presence of vortex shedding in the vicinity. Due to the integral design of the NPM, the RPV contains very few penetrations along the primary coolant flow path. For the NPM design, the only components that meet this criterion are the primary coolant flow sensors and RRVs, which are both located in the downcomer. Due to the flow conditions and geometry in these regions, no FIV mechanisms other than AR are credible for component and instrument ports.

2.3.6.6 ECCS Valves During an accident the Emergency Core Cooling System (ECCS) may be actuated, allowing flow through the ECCS valves. The ECCS valves are an angle globe valve design, wherein flow makes a 90 degree turn and passes around the valve disc when

© Copyright 2018 by NuScale Power, LLC 37

NuScale Comprehensive Vibration Assessment Program Technical Report TR-0716-50439-NP Draft Rev. 12 the valve is open (See Figure 2-25). The valve disc is not in direct cross flow, and downstream structures (the valve 90 degree turn) are present to disrupt any potential vortices generated by the valve internals. The valve body is designed for reaction loads of valve discharge and seismic loads, and is therefore thick-walled relative to schedule 160 piping. It is not a bounding component for turbulent buffeting analysis (Section 3.2.3). Due to the geometry in the ECCS valves, no FIV mechanisms are credible for through-valve flow.

Figure 2-25 ECCS Valve Internal Flow Diagram 2.4 Regulatory Requirements Consistent with RG 1.20, Section 2, the prototype CVAP for the NPM is composed of three sub-programs. The program includes

• a vibration and stress analysis program

• a vibration measurement program

• an inspection program The analysis program uses theoretical analysis to predict the natural frequencies, mode shapes, and structural responses of the NPM components to various sources of flow excitations.

The measurement program consists of prototype testing that is used to validate the analysis program inputs, results, and margins of safety. Prototype testing consists of separate effects, factory, and initial startup tests. The measurement program verifies the structural integrity of the NPM components. If discrepancies are identified between the analysis and the measurement programs, reconciliation is performed.

The inspection program consists of inspections of the applicable NPM components before and after initial startup testing in order to confirm that the vibratory behavior of the susceptible components is acceptable. Inspection is generally performed outside the

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