ML20050D019
| ML20050D019 | |
| Person / Time | |
|---|---|
| Issue date: | 02/19/2020 |
| From: | Walter Kirchner Advisory Committee on Reactor Safeguards |
| To: | Matthew Sunseri Advisory Committee on Reactor Safeguards |
| Snodderly, M, ACRS | |
| Shared Package | |
| ML20044D595 | List: |
| References | |
| Download: ML20050D019 (4) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, DC 20555 - 0001 February 19, 2020 MEMORANDUM TO: Matthew W. Sunseri, Chairman Advisory Committee on Reactor Safeguards FROM: Walter L. Kirchner, Chairman /RA/
NuScale Subcommittee Advisory Committee on Reactor Safeguards
SUBJECT:
PROPOSED RECOMMENDATION FOR ACRS REVIEW OF NUSCALE POWER, LLC, DESIGN CERTIFICATION APPLICATION - SAFETY EVALUATION WITH NO OPEN ITEMS FOR CHAPTER 1, INTRODUCTION AND GENERAL DESCRIPTION In response to the Committees request, I have reviewed the NRC staffs safety evaluation report (SER) with no open items for Chapter 1, Introduction and General Description, dated December 9, 2019 (ML20013D437). The following is my recommended course of action concerning further review of this chapter of the design certification application and the staffs associated safety evaluation.
SER Phase 4 Summary Chapter 1 of the SER documents the staffs review of Revision 3 of Chapter 1, Introduction and General Description of the Plant, of the NuScale Design Certification Application, Part 2, Tier 2 Final Safety Analysis Report (ML19241A398). The SER summarizes the staffs safety review of the NuScale design against the requirements of Title 10 of the Code of Federal Regulations (10 CFR) Part 52, Subpart B.
Chapter 1 of the applicants Final Safety Analysis Report is a general introduction to the NuScale design, including design bases and limits of operation; analysis of structures, systems, and components and the facility as a whole; and the information prescribed in 10 CFR 52.47(a) to be provided as part of standard design certification.
The design consists of up to 12 NuScale Power Modules (NPMs) in a single reactor building.
The NPMs are largely immersed in a large pool of water, which also serves as the ultimate heat sink. Each NPM is composed of a reactor core, a pressurizer, two helical-coil steam generators within a reactor pressure vessel, and housed integral to a compact steel containment vessel.
Each NPM is rated at 160 MWt, with approximately 50 MWe output.
M. Sunseri The staffs evaluation also provides additional information, including an index to the applicants 17 exemption requests (Section 1.14). In conclusion, the staff completed its review of Revision 3 of the Design Certification Application (DCA), as documented in the applicants Final Safety Analysis Report, and found it acceptable.
Applicable Concerns from ACRS Phase 3 Letter Report The committee did not conduct a prior interim review of Chapter 1 of the SER.
Staff Response to ACRS Letter Report Not applicable.
Open Items from Phase 3 Requiring Further ACRS Review Electrical Power: Specific to the Chapter 1 content, Section 1.4.3.2 addresses the applicability of Topical Report TR-0815-16497-P-A, Safety Classification of Passive Nuclear Power Plant Electrical Systems, a topic we reviewed in our letter report of July 26, 2017 (ML17205A380) and in our interim letter report on Chapters 7 and 8 of September 26, 2018 (ML18270A374). At issue was the significant open item of whether the DCA meets the specifications of that topical report and its associated safety evaluation limitations and conditions. The staff agreed with our conclusion and recommendation regarding Chapter 8, but disagreed on our additional recommendation of applying design, qualification, and quality assurance provisions to any non-safety alternating current or direct current power supplies that support operation of risk-significant systems or components, or performance of risk-significant human actions that are identified in the site-specific probabilistic risk assessment. The staff believes that this additional provision would be covered in the design certification applicants reliability assurance program.
In this Section 1.4.3.2, the staff discusses its disposition of each of the five limitations and conditions and conditions of applicability in TR-0815-16497-P-A, as applied to the highly reliable DC power system (EDSS).
This reviewer draws attention to Condition 4.4: Evaluate the frequency for which a combination of an anticipated operational occurrence (AOO) and an actuation of the NuScale emergency core cooling system is realistically expected to occur, and show that such a combination of events is not expected to occur during the lifetime of the module. The staff dispositioned this based on the applicants realistic analysis of such events (in contrast to Chapter 15 assumptions), which showed emergency core cooling system actuation is expected to occur much less than once in the lifetime of an NPM. However, in Section 6.2.1.1.4.1.5.3 Evaluation of Test Data, The NuScale DCA identifies an inadvertent opening of reactor recirculation valve opening (a credible AOO and not a loss-of-coolant accident) as the most limiting design basis event for calculated peak containment pressure By definition, an AOO is an event expected one or more times over the life of a plant. Further, there are 12 modules supported by a single EDSS. The staff should further clarify how this condition was met, or resolve this discrepancy.
Regarding Condition 4.5: Demonstrate that the reactor can be brought to a safe shutdown using only safety-related equipment in the absence of electrical power following a design basis event, with margin for stuck rods. Alternately, an applicant addressing this condition may provide justification, for NRC review, for a less restrictive approach. Section 15.0.6 evaluates a scenario in which a control rod assembly fails to insert and the chemical and volume control
M. Sunseri system is unavailable, resulting in the reactor core returning to a low power level. This relates to the exemption request to GDC 27. The staff goes on to assert the NuScale design can maintain a safe, stable condition with a stuck rod in the long term following an accident, For this reviewer, an uncontrolled return to criticality and power does not represent a stable condition, and is safe only to the extent that it can be demonstrated that specified acceptable fuel design limits and reactor coolant pressure boundary limits are not exceeded. This needs to be addressed as part of the Chapter 15 review and in our Boron Dilution and Return to Criticality Focus Area reviews.
Under Conditions of Applicability, Condition II.2 (Implement a methodology similar to the following to compare the reliability of the highly reliable DC system to that of a typical 1E DC power system), Section 8.3.2.1.1 states that an evaluation of EDSS reliability was performed We learned that an internal engineering report on EDSS reliability was reviewed by staff. We had not been previously provided this information, or report.
An ACRS Subcommittee meeting was held on February 4, 2020, to review the staffs evaluation of the EDSS against the TR. The staff explained how the EDSS meets each of the conditions of applicability listed in Table 3-1 and augmented design, qualification, and quality assurance provisions of Table 3-2 of the TR. The staffs SER found that the NuScale system meets the requirements of the TR and is acceptable.
Steam Generator: Section 1.7 addresses requirements for additional technical information, specifically testing programs. Sections 1.7.1.3 and 4 address the helical coil steam generator thermal-hydraulic performance testing with both an electrically heated test rig and a larger bundle in a fluid-heated test stand. The results appear inconclusive with regard to density wave oscillations, thermal fatigue, vibration and wear, and general operational stability (contrast to control rod drive tests, Sections 1.7.1.6 and 7, which provide more conclusive results). These matters were discussed with staff and applicant at a subcommittee meeting on February 4, 2020.
Recommendation As lead reviewer for NuScale Chapter 1, I recommend that the Committee not perform any additional Phase 5 review of this chapter, contingent on completing our review of steam generator performance as part of the Steam Generator Focus Area.
M. Sunseri February 19, 2020
SUBJECT:
PROPOSED RECOMMENDATION FOR ACRS REVIEW OF NUSCALE POWER, LLC, DESIGN CERTIFICATION APPLICATION - SAFETY EVALUATION WITH NO OPEN ITEMS FOR CHAPTER 1, INTRODUCTION AND GENERAL DESCRIPTION Package No.: ML20044D595 Accession No: ML20050D019 Publicly Available Y Sensitive N Viewing Rights: NRC Users or ACRS Only or See Restricted distribution *via email OFFICE ACRS/TSB SUNSI Review ACRS/TSB ACRS NAME MSnodderly MSnodderly LBurkhart (MSnodderly for) WKirchner (MSnodderly for)
DATE 2/19/2020 2/19/2020 2/19/2020 2/19/2020 OFFICIAL RECORD COPY