ML23038A030
| ML23038A030 | |
| Person / Time | |
|---|---|
| Site: | 99902037 |
| Issue date: | 02/14/2023 |
| From: | Siva Lingam Licensing Processes Branch |
| To: | Olinski D Westinghouse |
| References | |
| EPID L-2022-TOP-0043 | |
| Download: ML23038A030 (1) | |
Text
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION ENCLOSURE 2 REQUEST FOR ADDITIONAL INFORMATION RELATED TO PRESSURIZED WATER REACTOR OWNERS GROUP TOPICAL REPORT PWROG-21001-P/NP, REVISION 0, HYDROGEN-BASED TRANSIENT CLADDING STRAIN LIMIT DOCKET NO. 99902037 EPID NO. L-2022-TOP-0043 (NON-PROPRIETARY)
Proprietary information pursuant to Section 2.390 of Title 10 of the Code of Federal Regulations has been redacted from this document.
Redacted information is identified by blank space enclosed within double brackets.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION REQUEST FOR ADDITIONAL INFORMATION TO SUPPORT REVIEW OF PRESSURIZED WATER REACTOR OWNERS GROUP TOPICAL REPORT PWROG-21001-P/NP, REVISION 0, HYDROGEN-BASED TRANSIENT CLADDING STRAIN LIMIT DOCKET NO. 99902037 EPID NO. L-2022-TOP-0043
1.0 BACKGROUND
The U.S. Nuclear Regulatory Commission (NRC) staff is currently engaged in a review of a Pressurized Water Reactor Owners Group (PWROG) submitted Topical Report (TR)
PWROG-21001-P/NP, Revision 0, Hydrogen-Based Transient Cladding Strain Limit. By letter dated July 28, 2022 (Agencywide Documents Access and Management System (ADAMS)
Accession No. ML22209A254), PWROG requested NRC review and approval for referencing in regulatory actions per NRC the TR program. The purpose of this TR is to establish a new, alternative fuel performance design limit for transient cladding strain. The TR will facilitate PWROG licensees use of a data-driven, performance-based design limit when performing transient cladding strain analyses. This is an alternative limit which, after NRC approval, can be used in lieu of the current one percent transient cladding strain limit established in Section 4.2, Fuel System Design, of NUREG-0800, Standard Review Plan [SRP] for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition (ML070740002).
This TR is applicable to Westinghouse Electric Company nuclear steam supply system (NSSS) and Combustion Engineering NSSS PWROG members that use Westinghouse Performance Analysis and Design Model (PAD5) fuel performance models for ZIRLO and/or Optimized ZIRLO' high performance fuel cladding material.
The NRC staff conducted a virtual regulatory audit on January 18, 2023, to increase review efficiency (ADAMS Accession No. ML22361A156). After reviewing the TR and conducting the regulatory audit, the NRC staff has determined that the NRC needs the following additional information as noted in Section 3.0 of this document to complete its review of the TR.
2.0 REGULATORY BASES Regulatory guidance for the review of fuel system materials and designs and adherence to Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Appendix A, General Design Criteria [(GDC)] for Nuclear Power Plants, GDC-10, Reactor design, GDC-27, Combined reactivity control systems capability, and GDC-35, Emergency core cooling, is provided in the SRP, NUREG-0800, Section 4.2, Fuel System design. This SRP section provides regulatory guidance for the review of fuel rod cladding materials, the fuel system, and the design of the fuel assemblies.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION According to SRP Section 4.2, the fuel system safety review provides assurance that:
the fuel system is not damaged as a result of normal operation and anticipated operational occurrences (AOOs)
fuel system damage is never so severe as to prevent control rod insertion when it is
- required,
the number of fuel rod failures is not underestimated for postulated accidents, and
coolability is always maintained.
SRP Section 4.2, paragraph II.1.B.vi discusses the treatment of pellet-cladding interaction (PCI) and pellet-cladding mechanical interaction (PCMI) phenomena. Specifically, it states that no criterion currently exists for fuel failure resulting from PCI or PCMI, but that two related criteria should be applied: (1) the strain of the cladding during a transient should not exceed 1 percent; and (2) fuel melting should be avoided. Along with other acceptance criteria related to known fuel rod failure mechanisms, these criteria are used to meet the requirements of GDC-10, as it relates to specified acceptable fuel design limits for normal operation, including AOOs, and 10 CFR Part 100, Reactor Site Criteria, as it relates to fission product releases for postulated accidents. The PWROG is requesting to replace the 1 percent transient cladding strain (TCS) limit described in first criterion with a hydrogen-based TCS described in PWROG-21001-P for condition II overpower transients.
3.0 REQUEST FOR ADDITIONAL INFORMATION (RAI)
RAI-1
PCMI is established to be a biaxial phenomenon (i.e., the cladding is placed under both axial and radial stress). (( ))
RAI-2
In the proposed hydrogen-based transient cladding strain limit, (( ))
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION
RAI-3
It is stated on page 6-2 of Enclosure 1 to PWROG letter dated July 28,2022, that (( ))
RAI-4
When restarting after an AOO where cladding plastic strain was accumulated as a result of the AOO, how are the cladding strain hardening and possible opening of the fuel-cladding gap taken into account? In addition, how is further cladding strain measured or accounted for? Is the accumulated strain during the AOO subtracted from the proposed limit, such that the new transit cladding strain limit would be reduced for subsequent AOOs?
RAI-5
Please clarify which specific Condition II events the proposed TCS limit is planned to be applied to.
Principal Contributors: J. Messina, NRR P. Raynaud, RES Date: February 14, 2023