ML18319A380
| ML18319A380 | |
| Person / Time | |
|---|---|
| Site: | NuScale |
| Issue date: | 11/15/2018 |
| From: | Rad Z NuScale |
| To: | Document Control Desk, Office of New Reactors |
| References | |
| RAIO-1118-62971 | |
| Download: ML18319A380 (10) | |
Text
RAIO-1118-62971 November 15, 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" 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 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,
~ ~
/"Zackary W. Rad Director, Regulatory Affairs NuScale Power, LLC Distribution: Gregory Cranston, NRC, OWFN-8G9A Samuel Lee, NRC, OWFN-8G9A Marieliz Vera, NRC, OWFN-8G9A Enclosure 1: 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
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:
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.
NuScale classifies the design basis pipe break (DBPB) event, which includes spurious ECCS actuation, as a plant-wide service level C event. However, the service level for individual NSSS components may vary from the plant-wide designation based on the component function and allowable deformation during each event. It has been determined that ECCS actuation and the associated dynamic blowdown load service level classification should be changed for the ECCS valves because that is their primary function; therefore, service level A or B is appropriate. Level B was selected because the frequency of this event is low (five spurious openings, five actuations due to inadvertent RSV openings, and five possible actuations due to small break LOCAs) and because a small amount of deformation does not affect ECCS performance. As a level B service load, ECCS actuation is considered in fatigue evaluations.
Also during the clarification call, the NRC staff stated that a VT-1 examination of the ECCS valve flange bolts was unacceptable to use as the 10-year interval examination because of concerns related to the reliability of the VT-1 inspection. Therefore, NuScale will conduct UT examinations of these bolts once every 10 years whether or not they are removed during the 10-year interval.
Impact on DCA:
The FSAR Tier 2, Section 3.6.2 and Table 3.9-11 and Table 5.2-6 have been revised as described in the response above and as shown in the markup provided in this response.
NuScale Nonproprietary
Protection against Dynamic Effects Associated with Postulated Rupture NuScale Final Safety Analysis Report of Piping chosen because the DHRS cannot be isolated from the CNV as there are no isolation valves.
RAI 03.06.02-6 Breaks are not postulated in this segment of piping because it meets the design criteria for break exclusion in a containment penetration area (see Section 3.6.2.1.2). Although the DHRS condenser is manufactured from piping products, it is considered a major component and not a piping system, thus breaks are not postulated.
RAI 03.06.02-6, RAI 03.06.02-17, RAI 03.06.02-17S2 Connection of Reactor Vent Valves and Reactor Recirculation Valves to the Reactor Vessel In the NuScale design, each of three RVVs and each of two reactor recirculation valves bolt directly to reactor vessel nozzles. These five bolted-flange connections are classified as break exclusion areas. Because this break exclusion area does not include a physical piping length, a majority of the BTP 3-4 B.A.(ii) criteria do not apply. However, similar to the augmented ISI requirements given for piping welds in BTP 3-4 B.A.(ii),
augmented ISI requirements are specified for the bolts of these flanged connections to ensure they are inspected at least once per interval using a UT examination (Section 3.13.2).
3.6.3 Leak-Before-Break Evaluation Procedures RAI 03.06.02-6 General Design Criterion 4 includes a provision that the dynamic effects associated with postulated pipe ruptures may be excluded from the design basis when analyses reviewed and approved by the Commission demonstrate that the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping.
This analysis is called LBB. The LBB concept is based on the plant's ability to detect a leak in the piping components well before the onset of unstable crack growth.
For the NuScale Power Plant, the application of LBB is limited to the ASME Class 2 main steam and feedwater piping systems inside the CNV. The FWS piping analysis addresses significant feedwater cyclic transients and produces bounding loads for the ASME Class 2 piping with respect to LBB.
The methods and criteria to evaluate LBB are consistent with the guidance in Standard Review Plan 3.6.3 and NUREG-1061, Volume 3. Potential degradation mechanisms are described in Section 3.6.3.1; analysis for main steam and feedwater piping is provided in Section 3.6.3.4. Leak detection is discussed in Section 3.6.3.5.
3.6.3.1 Potential Degradation Mechanisms for Piping In high-energy piping systems, environmental and operating material degradation could adversely affect the integrity of the system as well as the piping system LBB applicability. The application of LBB requires that the affected systems not be susceptible to environmental and operating degradation mechanisms such as erosion/
Tier 2 3.6-37 Draft Revision 3
NuScale Final Safety Analysis Report Mechanical Systems and Components RAI 03.06.02-17S2 Table 3.9-11: Load Combinations for Emergency Core Cooling System Valves Plant Event Service Level Load Combination(1) Allowable Limit Design Design P + DW + B + EXT Design Testing Testing P + DW + B + EXT Testing Normal operation A P + DW + B + EXT + TH Level A Transients B P + DW + B + EXT + TH Level B Transients + OBE(2)(3) B P + DW + B + EXT + TH +/- OBE Level B Design basis pipe breaks B(7) P + DW + B + EXT + DBPB Level B(7)
C(8) Level C(8)
Hydrogen detonation C Pg1 + DW + B Level C SG tube failure(4) C P + DW + B + EXT + R Level C Rod ejection accident D P + DW + B + EXT + REA Level C(5)
Main steam and feedwater pipe D P + DW + B + EXT + MSPB/FWPB Level D breaks SSE + DBPB/MSPB/FWPB D P + DW + B + EXT +/- SRSS(SSE + DBPB/ Level D MSPB/FWPB)(6)
Hydrogen DDT D Pg2 + DW + B Level D Notes:
- 1. Applicable loads are defined in Section 3.9.3.1.1 and Table 3.9-2.
- 2. Fatigue analysis of all applicable components is evaluated in accordance with the applicable ASME Code considering the effects of the PWR environment in accordance with RG 1.207.
- 3. OBE is included in fatigue analyses.
- 4. Dynamic load due to SG tube failure is negligible.
- 5. In accordance with NUREG-0800 Section 15.4.8, Acceptance Criterion 2.
- 6. Dynamic loads are combined considering the time phasing of the events in accordance with RG 1.92 and NUREG-0484.
- 7. RRV and RVV actuation loads, including the dynamic discharge thrust load, are considered service level B loads for the ECCS valves. These loads are to be considered in fatigue evaluations.
- 8. DBPB loads other than RRV and RVV actuation are considered service level C.
Tier 2 3.9-78 Draft Revision 3
NuScale Final Safety Analysis Report Integrity of Reactor Coolant Boundary RAI 03.06.02-17, RAI 03.06.02-17S2, RAI 05.02.04-3, RAI 05.03.01-3, RAI 05.04.02.01-6, RAI 06.06-3 Table 5.2-6: Reactor Pressure Vessel Inspection Elements Description Examination Examination Notes Category Method RPV Shell and Head Welds Lower RPV flange shell to RPV bottom head B-A Volumetric Upper RPV flanged transition shell to lower SG shell Lower SG shell to upper SG shell Upper SG shell to integral steam plenum Integral steam plenum to PZR shell PZR shell to RPV top head Steam plenum cap to integral steam plenum RPV Internal Welds Core support block to RPV bottom head B-N-2 VT-3 Core support block to latch Core barrel guide to lower RPV flange shell Upper SG support to lower RPV integral steam plenum Lower SG support to upper RPV Instrumentation and Controls Sleeve Welds None None These welds are part of the cladding.
Flow diverter to RPV lower head B-N-1 VT-3 B-N-1 is for the space above and RPV interior surfaces and attachment welds below the core made accessible by removal of components during a normal refueling outage RPV External Welds RPV support plate to RPV support gussets F-A VT-3 RPV support plate to upper RPV SG shell 1-4 RPV support plate to upper RPV SG shell B-K Surface or RPV support gussets to upper RPV SG shell Volumetric RPV lateral support lug RPV Nozzle to Shell and Head Welds Reactor recirc valve flange B-D Volumetric Inside corner. All welds Feedwater nozzles examination requirement IWB-2500-7(d).
RCS discharge Main steam nozzles B-D Volumetric Examination requirement IWB-2500-7(d)
RCS injection B-D N/A No inside corner PZR spray supply lines Tier 2 5.2-37 Draft Revision 3
NuScale Final Safety Analysis Report Integrity of Reactor Coolant Boundary Table 5.2-6: Reactor Pressure Vessel Inspection Elements (Continued)
Description Examination Examination Notes Category Method Bolting RPV main flange bolts B-G-1 See note Per Note 1 of B-G-1, surface examination is permitted when bolts are removed.
RVV and RRV flange threaded fasteners B-G-2, augmented VT-1Volumetric This inspection is required to be completed once every inspection interval. If the connection is not planned to be removed during the interval, a volumetric exam is required to be completed at least once per interval.Augmented inspection to follow the guidance of B-G-1.
RPV bolting two inches or less in diameter B-G-2 VT-1 Examined if removed.
Assembled RPV RPV- assembled after refueling outage B-P VT-2 Per Section XI IWA-5241(c),
leakage is continuously monitored in the CNV and constitutes a VT-2 examination.
Tier 2 5.2-39 Draft Revision 3