ML25013A241
| ML25013A241 | |
| Person / Time | |
|---|---|
| Site: | 05200050 |
| Issue date: | 01/13/2025 |
| From: | NuScale |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML25013A204 | List: |
| References | |
| LO-177590, LO-178078 | |
| Download: ML25013A241 (1) | |
Text
Response to SDAA Audit Question Question Number: DWO-SC-37 Receipt Date: 11/18/2024 Question:
In the response to DWO-SC-23, FSAR page 5.4-13 is included as a markup. In Section 5.4.1.6.1 the markup contains the verbiage For the first NPM only NRC requests that this language be clarified to mean the first NPM put in service, or the first NPM subject to inspection.
Response
The attached markup to Chapter 1 and Chapter 5 of the Final Safety Analysis Report and to the Technical Specifications incorporates more specific language outlining the inspection requirements for the first NuScale Power Module (NPM-20) to undergo a refueling outage.
Markups of the affected changes, as described in the response, are provided below:
NuScale Nonproprietary NuScale Nonproprietary
NuScale Final Safety Analysis Report Interfaces with Standard Design NuScale US460 SDAA 1.8-8 Draft Revision 2 Audit Question A-5.PTLR-6 Audit Question A-5.PTLR-9 Audit Question A-5.PTLR-10 COL Item 5.3-1:
An applicant that references the NuScale Power Plant US460 standard design will choose the final transients to generate the reactor vessel pressure-temperature limits report and limiting conditions for operation. An applicant that references the NuScale Power Plant US460 standard design will confirm that the design geometries, final transients, and material properties of the reactor pressure vessel are bounded by (or identical to) those used in the Pressure and Temperature Limits Methodology to confirm that the example curves in the Standard Design Approval Application are applicable.
Operating procedures will ensure that pressure-temperature limits for the as-built reactor are not exceeded. These procedures will be based on the limits defined in the pressure-temperature limits report and material properties of the as-built reactor vessel.An applicant that references the NuScale Power Plant US460 standard design will develop operating procedures to ensure that transients will not be more severe than those for which the reactor design adequacy had been demonstrated. These procedures will be based on material properties of the as-built reactor vessels.
5.3 Audit Question A-3.9.2-26-F, Audit Question A-3.9.2-28, Audit Question DWO-SC-23, Audit Question DWO-SC-24, Audit Question DWO-SC-36, Audit Question DWO-SC-37 COL Item 5.4-1:
An applicant that references the NuScale Power Plant US460 standard design will develop and implement a Steam Generator Program for periodic monitoring of the degradation of steam generator components to ensure that steam generator tube integrity is maintained. The Steam Generator Program will be based on the latest revision of Nuclear Energy Institute NEI 97-06, Steam Generator Program Guidelines, and applicable Electric Power Research Institute steam generator guidelines at the time of the application. The elements of the program will include: assessment of degradation, tube inspection requirements, tube integrity assessment, tube plugging, primary-to-secondary leakage monitoring, shell side integrity assessment, primary and secondary side water chemistry control, foreign material exclusion, loose parts management, contractor oversight, self-assessment, and reporting.
The Steam Generator Program for the NuScale Power Plant US460 will require 100 percent tube inspections at the first refueling outage. In addition to other requirements and inspections, the Steam Generator Program for the first module to undergo a refueling outage will require at least 20 percent tube inspections during each subsequent refueling outage for the first 72 effective full power months after the first refueling outage.
Subsequent applicants that reference the NuScale Power Plant US460 design shall provide justification that the results of the first module's inspections are applicable to the subsequent modules in order to demonstrate that these additional inspection requirements are not applicable to the subsequent modules.
5.4 COL Item 6.2-1:
An applicant that references the NuScale Power Plant US460 standard design will verify that the final design of the containment vessel meets the design-basis requirement to maintain flange contact pressure at accident temperature, concurrent with peak accident pressure.
6.2 COL Item 6.3-1:
An applicant that references the NuScale Power Plant US460 standard design will describe a containment cleanliness program that limits debris within containment. The program should contain the following elements:
- Foreign material exclusion controls to limit the introduction of foreign material and debris sources into containment.
- Maintenance activity controls, including temporary changes, that confirm the emergency core cooling system function is not reduced by changes to analytical inputs or assumptions or other activities that could introduce debris or potential debris sources into containment.
- Controls that prohibit the introduction of coating materials into containment.
- An inspection program to confirm containment vessel cleanliness before closing for normal power operation.
6.3 COL Item 6.4-1:
An applicant that references the NuScale Power Plant US460 standard design will comply with Regulatory Guide 1.78 Revision 2, "Evaluating the Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release."
6.4 Table 1.8-1: Combined License Information Items (Continued)
Item No.
Description of COL Information Item Section
NuScale Final Safety Analysis Report Reactor Coolant System Component and Subsystem Design NuScale US460 SDAA 5.4-10 Draft Revision 2 current test meets Electric Power Research Institute (EPRI) 1013706 (Reference 5.4-2).
Preservice examinations performed in accordance with the ASME BPVC,Section III, Subsubarticle NB-5280 and Section XI, Subarticle IWB-2200 (Reference 5.4-5) use examination methods of ASME BPVC Section V, except as modified by Section III, Paragraph NB-5111. These preservice examinations include essentially 100 percent of the pressure boundary welds.
Audit Question A-5.4.1.4-1, Audit Question A-5.4.1.6.1-1, Audit Question DWO-SC-26 RAI 5.4.1.6.1-1 A preservice volumetric, full-length preservice inspection of essentially 100 percent of the tubing in each SG is performed. The length of the tube extends from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet. The tube-to-tubesheet welds are not part of the tube. The preservice inspection is performed after tube installation and shop or field primary-side hydrostatic testing and before initial power operation to provide a definitive baseline record, against which future ISI can be compared. Technical Specifications Section 5, Administrative Controls, defines the tube plugging criterion as the maximum allowable flaw in the tube wall. Tubes with flaws that are equal to or exceed 40 percent of the nominal tube wall thicknessthe tube plugging criterion are plugged. Tubes with flaws that could potentially compromise tube integrity before the performance of the first ISI, and tubes with indications that could affect future inspectability of the tube, are also plugged. The volumetric technique used for the preservice examination is capable of detecting the types of preservice flaws that may be present in the tubes and permits comparisons to the results of the ISI expected to be performed to satisfy the SG tube inspection requirements in accordance with the plant technical specifications.
Audit Question A-3.9.2-26-F, Audit Question A-3.9.2-28, Audit Question DWO-SC-23, Audit Question DWO-SC-24, Audit Question DWO-SC-36, Audit Question DWO-SC-37 As discussed above, the operational inservice testing and inspection programs described in Section 5.2.4, RCPB ISI and Testing, and the SG program described in Section 5.4.1.6, Steam Generator Program, provide testing and inspection requirements following initial plant startup. The SG inlet flow restrictors are examined by VT-3 in accordance with IWA-2213 when removed for SG tube examinations. Inservice inspection and testing of the SGS steam and feedwater piping is described in Section 6.6.
5.4.1.5 Steam Generator Materials Selection and fabrication of pressure boundary materials used in the SGs and associated components are in accordance with the requirements of ASME BPVC Section III and Section II as described in Section 5.2.3, RCPB Materials, and the materials used in the fabrication of the SGs are in Table 5.2-3.
The RCPB materials used in the SGS are Quality Group A and their design, fabrication, construction, tests, and inspections conform to Class 1 in accordance with the ASME BPVC and the applicable conditions promulgated in 10 CFR 50.55a(b). The SGS materials forming the RCPB, including weld materials, conform to fabrication, construction, and testing requirements of ASME
NuScale Final Safety Analysis Report Reactor Coolant System Component and Subsystem Design NuScale US460 SDAA 5.4-13 Draft Revision 2 degradation allowance (additional tube wall thickness above minimum required for design) as discussed in Section 5.4.1.2, System Design. The NPM reactor coolant flowrates are also lower than the flowrates across the SG tubes in PWR recirculating SGs as discussed in Section 5.1, RCS and Connecting Systems.
This low flow rate reduces the flow energy available to cause FIV wear degradation of SG tubes. Based on the additional tube wall margin and the additional margin against FIV turbulent buffeting wear (the most likely SG tube degradation mechanism), application of the existing PWR SG Program requirements to the design is appropriate.
For SGs in the PWR fleet with SB-163 UNS N06690 SG tubing, the only observed degradation has been wear as a result of FIV (tube-to-tube or tube-to-support plate) or wear due to foreign objects. With respect to the risk of introduction of foreign objects, the NPM is at no greater risk than existing designs; therefore, the design does not warrant deviations from existing SG program guidelines. From the standpoint of SG tube design, the two significant differences between the SG design and current large PWR designs is the helical shape of the SG tubing and the SG tube support structure. The helical shape of the SG tubing itself does not represent risk of degradation based on the minimum bend radius of the helical tubing being within the historical experience base of PWR SG designs. Prototypic testing of the SG tube supports validates acceptable performance (including wear) of the SG tube support design. Implementation of a typical SG program is appropriate based on evaluation of the design of the SG tube supports.
5.4.1.6.1 Degradation Assessment Audit Question A-5.4.1.6.1-1, Audit Question DWO-SC-26 RAI 5.4.1.6.1-1 A degradation assessment of the NPM SG identifies several potential degradation mechanisms. Wear is the most likely degradation mechanism, and there is the potential for several secondary side corrosion mechanisms, including under deposit pitting and intergranular attack based on the once-through design with secondary boiling occurring inside the tubes. The estimated growth rates for these potential defects is sufficiently low that the SG tube plugging criterion for the SG is a 40 percent through wall defect.
Operational SG tube integrity is ensured by implementing tube plugging criteria, implementing elements of the SG program, and implementing the SG inspections.
Audit Question A-3.9.2-26-F, Audit Question A-3.9.2-28, Audit Question DWO-SC-23, Audit Question DWO-SC-24, Audit Question DWO-SC-36, Audit Question DWO-SC-37 A 100 percent SG tube inspection is completed during the first refueling outage following initial startup or SG replacement. After the first refueling outage, a 100 percent SG inspection is completed on a staggered basis over the next 72 effective full power months in order to evaluate ongoing SG tube degradation.
COL Item 5.4-1:
An applicant that references the NuScale Power Plant US460 standard design will develop and implement a Steam Generator Program for periodic monitoring of the degradation of steam generator components to ensure that
NuScale Final Safety Analysis Report Reactor Coolant System Component and Subsystem Design NuScale US460 SDAA 5.4-14 Draft Revision 2 steam generator tube integrity is maintained. The Steam Generator Program will be based on the latest revision of Nuclear Energy Institute NEI 97-06, "Steam Generator Program Guidelines," and applicable Electric Power Research Institute steam generator guidelines at the time of the application.
The elements of the program will include: assessment of degradation, tube inspection requirements, tube integrity assessment, tube plugging, primary-to-secondary leakage monitoring, shell side integrity assessment, primary and secondary side water chemistry control, foreign material exclusion, loose parts management, contractor oversight, self-assessment, and reporting.
Audit Question A-3.9.2-26-F, Audit Question A-3.9.2-28, Audit Question DWO-SC-23, Audit Question DWO-SC-24, Audit Question DWO-SC-36, Audit Question DWO-SC-37 The Steam Generator Program for the NuScale Power Plant US460 will require 100 percent tube inspections at the first refueling outage. In addition to other requirements and inspections, the Steam Generator Program for the first module to undergo a refueling outage will require at least 20 percent tube inspections during each subsequent refueling outage for the first 72 effective full power months after the first refueling outage. Subsequent applicants that reference the NuScale Power Plant US460 design shall provide justification that the results of the first module's inspections are applicable to the subsequent modules in order to demonstrate that these additional inspection requirements are not applicable to the subsequent modules.
5.4.2 Reactor Coolant System Piping 5.4.2.1 Design Basis Pressure-retaining portions of piping that penetrate the RCS form, in part, the RCPB as defined in 10 CFR 50.2 and include the PZR spray supply, RCS injection, RCS discharge, and RPV high-point degasification piping.
5.4.2.2 Design Description Section 6.2, Containment Systems, describes how each of the RCS lines enter containment through nozzle safe ends on the containment upper head and contain two containment isolation valves mounted on the outside of the containment.
A single PZR spray supply line enters through the containment head. This line branches inside containment into two PZR spray supply lines, each welded to a nozzle safe end on the RPV upper head with a corresponding spray nozzle inside the RPV near the top of the PZR space.
The RPV high-point degasification line is a single line that connects the containment upper head to a nozzle safe end on the RPV upper head.
The RCS injection line connects the containment upper head to a nozzle safe end on the side of the RPV. Inside the RPV, the line continues from the RPV wall, through the lower portion of the upper riser assembly and terminates near the
Programs and Manuals 5.5 NuScale US460 5.5-5 Draft Revision 2 5.5 Programs and Manuals 5.5.4 Steam Generator (SG) Program (continued)
- 3.
The operational LEAKAGE performance criterion is specified in LCO 3.4.5, "RCS Operational LEAKAGE.
- c.
Provisions for SG tube plugging criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding [40%] of the nominal tube wall thickness shall be plugged.
- d.
Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube plugging criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. A degradation assessment shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.
- 1.
Inspect 100% of the tubes in each SG during the first refueling outage following initial startup or SG replacement.
- 2.
After the first refueling outage following SG installation, inspect 100% of the tubes in each SG at least every 7296 effective full power months, which defines the inspection period.
- 3.
If crack indications are found in any SG tube, then the next inspection for each affected and potentially affected unit SG for the degradation mechanism that caused the crack indication shall be at the next refueling outage. If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.
- e.
Provisions for monitoring operational primary to secondary LEAKAGE.