ML25091A117
| ML25091A117 | |
| Person / Time | |
|---|---|
| Issue date: | 04/21/2025 |
| From: | Harrington C Advisory Committee on Reactor Safeguards |
| To: | Walter Kirchner Advisory Committee on Reactor Safeguards |
| References | |
| Download: ML25091A117 (1) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, DC 20555 - 0001 April 21, 2025 MEMORANDUM TO:
Walter L. Kirchner, Chairman Advisory Committee on Reactor Safeguards FROM:
Craig Harrington, Member Advisory Committee on Reactor Safeguards
SUBJECT:
INPUT FOR ACRS REVIEW OF THE NUSCALE STANDARD DESIGN APPROVAL APPLICATION - SAFETY EVALUATION WITH NO OPEN ITEMS FOR CHAPTER 5, REACTOR COOLANT SYSTEM AND CONNECTING SYSTEMS In response to the Committees request, I have reviewed the NRC staffs safety evaluation report (SER) provided to support ACRS review of the standard design approval application (SDAA), with no open items for Chapter 5, Reactor Coolant System and Connecting Systems, dated October 31, 2023. The following is my recommended course of action concerning further review of this chapter of the SDAA and the staffs associated SER.
SER Summary Chapter 5 of the SER documents the staffs review of Revision 1 of Chapter 5, Reactor Coolant System and Connecting Systems, of the NuScale SDAA Final Safety Analysis Report (FSAR). Additionally, this advance SER also reflects updated text which will be incorporated into Revision 2 and docketed later.
The reactor coolant system (RCS) is a subsystem of the NuScale Power Module that provides for circulation of the primary coolant. The applicants design relies on natural circulation flow for the primary coolant and does not include reactor coolant pumps or an external piping system. The RCS includes the reactor vessel and integral pressurizer, the reactor vessel internals, the reactor safety valves, RCS piping inside the containment vessel (RCS injection and discharge, pressurizer spray supply, and reactor vessel high-point degasification lines),
pressurizer controls, and RCS instruments and cables. The two helical-coil steam generators (SGs) are contained within the reactor vessel and, as an interfacing system to the RCS, form a portion of the reactor coolant pressure boundary.
The staff has completed a thorough review of this chapter and has concluded that the NuScale design elements therein and associated combined license items met regulatory criteria.
Discussion Although the US460 design is fundamentally the same as the US600 approved in the DCA, notable changes have been made, including within both the reactor pressure vessel and
W.L. Kirchner integral SGs. This evolution to a more robust design as reflected in the US460 plant is a positive development, and in several instances, has effectively addressed areas of prior concern to the Committee.
The lower reactor pressure vessel shell in the US460 will be austenitic stainless steel rather than low alloy steel as planned for the certified design and as used within the legacy pressurized water reactor fleet. This change in material is discussed in detail in NuScale Technical Report TR-130721-P, Use of Austenitic Stainless Steel for NuScale Power Module Lower Reactor Pressure Vessel, which provides technical justification that the requirements in Title 10 of the Code of Federal Regulations (10 CFR) 50.60 (fracture toughness and material surveillance program requirements for the reactor coolant pressure boundary) and 10 CFR 50.61 (protection against pressurized thermal shock events) do not apply. The applicant has requested exemptions to these requirements which the staff has found to be acceptable.
NuScale has continued to evolve their understanding of density wave oscillation (DWO) and its potential impact on operation of the helical coil SGs through testing and analysis, leading to adjustments reflected in the US460 design. The SG inlet flow restrictor has been redesigned from that of US600 and now will be installed in each SG tube inlet to impose a suitable pressure drop for avoiding DWO within the normal operational power range. DWO conditions may still be encountered during startup and low power operations, resulting in a slow accumulation of SG tube damage. Previously, the applicant was focused on demonstration via testing that DWO conditions challenging to system components and operation could be avoided. In the latest revisions to the US460 FSAR, a DWO management strategy based on an Approach Temperature has been established. Although the associated FSAR revisions have not been formally reviewed by the Committee, they have been reviewed by the staff and the applicant presented the methodology and technical basis in detail during our February 4, 2025, subcommittee meeting. The Approach Temperature, defined as the difference between RCS Thot and Main Steam outlet temperature, is directly correlated to DWO margin. Applicant has further developed an Approach Temperature limit, below which DWO onset could occur.
Under the planned DWO management strategy, cumulative time in conditions favorable to DWO is tracked against a Technical Specification limit, in combination with SG tube inspections, to ensure the module remains well removed from unacceptable DWO-related damage accumulation.
The staffs review of these and other changes reflected in the US460 design, as compared to the US600 certified design, was thorough with well supported conclusions throughout.
Member questions and concerns were addressed by both the applicant and staff during the February 4, 2025, subcommittee meeting on Chapter 5. However, there are certain lines of questioning that were deferred to the remaining FSAR Chapter and Technical Report reviews, including Chapters 15 and 19.
Recommendation As lead reviewer for NuScale Chapter 5, I recommend that the Committee not perform any additional review of this chapter.
References
- 1. U. S. Nuclear Regulatory Commission, Safety Evaluation of NuScale SDAA Chapter 5, Reactor Coolant System and Connecting Systems, January 29, 2025 (ADAMS Accession Nos. ML24330A280 (Public) ML25028A418 (Non-Public)).
W.L. Kirchner 2. NuScale Power, LLC, Standard Design Approval Application, Part 2, Chapter 5, Reactor Coolant System and Connecting Systems, Revision 1, October 31, 2023 (ADAMS Accession No. ML23304A338).
- 3. U. S. Nuclear Regulatory Commission, Safety Evaluation of NuScale Pressure and Temperature Limits Methodology, January 29, 2025 (ADAMS Accession No. ML24332A008).
- 4. NuScale Power, LLC, Technical Report TR-130877, Pressure and Temperature Limits Methodology, December 31, 2022 (ADAMS Accession Nos. ML23304A341 (Public) and ML23304A342 (Non-Public)).
- 5.
NuScale Power, LLC, Technical Report TR-130721-P Use of Austenitic Stainless Steel for NuScale Power Module Lower Reactor Pressure Vessel, December 31, 2022 (ADAMS Accession Nos. ML23304A343 (Public) and ML23304A344 (Non-Public)).
W.L. Kirchner April 21, 2025
SUBJECT:
INPUT FOR ACRS REVIEW OF THE NUSCALE STANDARD DESIGN APPROVAL APPLICATION - SAFETY EVALUATION WITH NO OPEN ITEMS FOR CHAPTER 5, REACTOR COOLANT SYSTEM AND CONNECTING SYSTEMS Package Accession No: ML25091A091 Accession No: ML25091A117 Publicly Available (Y/N): Y Sensitive (Y/N): N If Sensitive, which category?
Viewing Rights:
NRC Users or ACRS only or See restricted distribution OFFICE ACRS SUNSI Review ACRS ACRS NAME MSnodderly MSnodderly LBurkhart CHarrington DATE 3/31/25 3/31/25 3/31/25 4/10/25 OFFICIAL RECORD COPY