ML19310F085
| ML19310F085 | |
| Person / Time | |
|---|---|
| Issue date: | 11/19/2019 |
| From: | Jordan Hoellman NRC/NRR/DANU/UARP |
| To: | John Segala NRC/NRR/DANU/UARP |
| Hoellman J, NRR/DANU/UARP, 301-415-5481 | |
| References | |
| Download: ML19310F085 (7) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 November 19, 2019 MEMORANDUM TO:
John P. Segala, Chief Advanced Reactor Policy Branch Division of Advanced Reactors and Non-Power Production and Utilization Facilities Office of Nuclear Reactor Regulation FROM:
Jordan P. Hoellman, Project Manager /RA/
Advanced Reactor Policy Branch Division of Advanced Reactors and Non-Power Production and Utilization Facilities Office of Nuclear Reactor Regulation
SUBJECT:
AUDIT REPORT OF REGULATORY AUDIT RELATED TO EPRI TOPICAL REPORT, URANIUM OXYCARBIDE (UCO)
TRISTRUCTURAL ISOTROPIC (TRISO) COATED PARTICLE FUEL PERFORMANCE The U.S. Nuclear Regulatory Commission staff conducted an audit of documents related to the Electric Power Research Institute (EPRI) Topical Report EPRI-AR-1, Uranium Oxycarbide (UCO) Tristructural Isotropic (TRISO) Coated Particle Fuel Performance (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML19155A173). The audit was conducted at Idaho National Laboratory, Idaho Falls, Idaho on October 8 and 9, 2019.
The plan for this audit is available in ADAMS under Accession No. ML19268A925. A summary of the audit is enclosed.
CONTACT: Jordan Hoellman, NRR/DANU 301-415-5481
Enclosure:
As stated
- via email NRC-008 OFFICE NRR/DANU/UARP:PM
- NRR/DANU/UART:TR
- NRR/DANU/UART:BC
- NRR/DANU/UARP:BC NAME JHoellman BTravis MHayes JSegala DATE 11/19/2019 11/07/2019 11/17/2019 11/19/2019
Enclosure REPORT OF REGULATORY AUDIT RELATED TO TOPICAL REPORT Uranium Oxycarbide (UCO) Tristructural Isotropic (TRISO) Coated Particle Fuel Performance A.
Background and Purpose By letter dated May 31, 2019, the Electric Power Research Institute (EPRI) submitted for U.S.
Nuclear Regulatory Commission (NRC) staff review, Uranium Oxycarbide (UCO) Tristructural Isotropic (TRISO) Coated Particle Fuel Performance, Topical Report EPRI-AR-1 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML19155A173). During the course of the technical review, NRC staff identified areas where additional information and detail are needed to make a safety finding. The purpose of the audit is to review the relevant reports, calculations, and sources to verify the information and conclusions in the topical report.
B.
Bases This regulatory audit is based on the following:
10 CFR 50.34(a)(3)(i), which requires in part that an applicant for a construction permit to build a power reactor provide principal design criteria for the facility. Similar regulatory requirements exist for design certification, combined license, and standard design approvals (10 CFR 52.47(a)(3)(i), 10 CFR 52.79(a)(4)(i), and 10 CFR 52.137(a)(3)(i),
respectively).
General Design Criterion (GDC) 10, Reactor design, which requires that the reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.
10 CFR 50.34(a)(1)(ii)(C), which requires an applicant describe the extent to which the reactor incorporates unique, unusual or enhanced safety features having a significant bearing on the probability or consequences of accidental release of radioactive materials.
The audit plan is available in ADAMS under Accession Number ML19268A925.
C.
Audit Location and Dates Location: The audit was conducted at Idaho National Laboratory, Idaho Falls, Idaho.
Date: October 8 and 9, 2019.
D.
Audit Team Members Jeff Schmidt, Senior Reactor Systems Engineer (Advanced Reactor Technical Branch)
Chris Van Wert, Senior Reactor Systems Engineer (Advanced Reactor Technical Branch)
Boyce Travis, Reactor Systems Engineer (Advanced Reactor Technical Branch), Audit Lead Antonio Barrett, Reactor Systems Engineer (Advanced Reactor Technical Branch)
Jordan Hoellman, Project Manager (Advanced Reactor Policy Branch)
E.
Applicant and Industry Participants Andrew Sowder *, EPRI Cristian Marciulescu, EPRI Steven Nesbit, LMNT Consulting (EPRI contractor)
Farshid Shahrokhi, Framatome Jim Tomkins, Kairos Power Steve Schilthelm, BWXT Mike Davenport, Idaho National Laboratory (INL)
Paul Demkowicz, INL Jess Gehin, INL David Laug, INL Doug Marshall, INL Michelle Sharp, INL John Stempien, INL Gerhard Strydom, INL
- Participated remotely F.
Documents Audited During the course of the audit, staff reviewed the following documents:
NE-20, Advanced Gas Reactor Fuel Development and Qualification Program - Quality Assurance Requirement and Interface Description, Rev. 0.
PLN-1468, Quality Assurance Program Plan for AGR Fuel Development and Qualification Program, Rev. 0.
EDF-4542, Statistical Sampling Plan for AGR Fuel Materials, Rev. 7.
EGG-NPR-10130, Evaluation of NP-MHTGR Performance Test Fuel Quality Control Data, dated February 1992.
G.
Description of Audit Activities and Summary of Observations As part of the audit, NRC staff held discussions primarily centered around the topics listed in the audit plan:
- 1. Advanced Gas Reactor (AGR) post-irradiation examination (PIE) that is referenced in the topical report;
- 2. The particle characterization dataset development leading to the topical report Table 5-3;
- 3. Further detail related to data used to develop topical report Table 5-5;
- 4. Temperature margin test data sets discussed in the topical report;
- 5. Context on the quality assurance (QA) program used in data acquisition for the data used in the topical report;
- 6. Additional information related to strontium, europium, and silver releases during normal operation (as defined in the topical report);
- 7. The role of the topical report in defining acceptable peak accident conditions as opposed to any transient conditions (e.g. rapid temperature changes);
- 8. The role of the topical report in defining specified acceptable system radionuclide release design limits (SARRDLs); and
- 9. The planned use of future AGR data to support the conclusions referenced in the report.
Initial discussion focused on the broader scope areas in topics 7, 8 and 9 above. This discussion emphasized the limited scope of the report as a partial fuel qualification document.
Any vendor referencing the report would have to address applicability of the concepts in the report to their specific fuel type. In areas where data is not provided or may not fully apply to the design type (such as accident conditions), the staff would expect the reactor designer to provide supplemental information for their specific fuel type. Along similar lines, the applicant stated SARRDLs were a function of the overall design philosophy and interaction with other design-specific systems, and so while the TRISO particle would play some role in the retention and release and radionuclides, the extent of the reliance on the particle would be up to the specific reactor designer. The applicant noted that future AGR data is not and will not be relied on to support the conclusions in the topical report.
Discussions with applicant and industry participants indicated that in general, the scope of the report conclusions was limited to the fuel particle itself. Thus, the end user or reactor designer would need to provide information on areas related to the fuel form (i.e. compact, pebble, other) such as radionuclide retention by the larger fuel form, contamination by trace uranium, the impact of the fuel/matrix compacting process, and others, and this topical report would be a single element of a larger fuel qualification effort. The applicant noted that the data on the compacts used in AGR and included in the topical report could be useful or provide insights for reactor designers but confirmed that the applicability of the conclusions in the report is limited to the fuel particle boundary.
In discussions on the quality assurance (QA) program used to collect the data, technical staff from INL provided context on the INL QA program, including providing the above referenced documents (NE-20 and PLN-1468) related to the quality assurance of the data collection. The applicant further provided references to NRC letters related to the review of the Next Generation Nuclear Plant (NGNP) QA program, which governed the testing and data acquisition for the activities described in the topical report (see ADAMS Accession No. ML12241A157). The applicant explained that data was captured under an NQA-1 compliant program, subject to audit by DOE; however, this was not explained in the topical report. As part of reviewing the data, staff also examined document EDF-4542 noted above; the design file for coated particle data, and sampling matched the values reported in the topical report.
One focus area for staff during the audit was clearly defining the performance criteria under which the report would apply. The conclusion of the report mentions temperature and burnup, though other factors (such as fluence) are referred to in the report. Time at temperature and the margin testing performed at higher temperatures would also provide applicable data for applicants referencing the report. Discussions between NRC staff and applicant and industry participants indicated a consolidated list of performance bounds to be referenced in the conclusion might be possible.
Based on discussions with the applicant and the documents reviewed, one area that staff did not find fully explored in the report was information related to short lived fission products that would not be detected by the gaseous monitor during irradiation testing of the TRISO particles.
Due to discrepancy between the residence time before examination and the half-life of some short-lived fission products, no data on a small subset of radionuclides is available. Applicant and industry participants stated historical data indicates these radionuclides would be characterized in a similar fashion to fission gases, which has available data. Because the topical report is of limited scope (applying to the particles and collected AGR data), this may not be a concern; however, staff does expect that a future applicant referencing the topical report would provide a justification for whatever approach is used to characterize the transport of all fission products.
Staff also pointed out areas where Table 5-5, Particle layer property 95% confidence values on means and dispersion limits, might not cover all the elements of the fuel specification based on the information provided in the report. Some potential missing elements included the carbon content of the fuel (based on the requested applicability of UCO fuel only), the use of an uninterrupted coating process, key similarities to the AGR fuel that are not captured in the table such as kernel to buffer ratio or relative particle size, and columnar grain structure of the silicon carbide (SiC) layer. NRC staff understood based on the discussion some of these parameters might be inherently constrained by the values provided in Table 5-5, but not all of them. Further, any derived constraint from the table was not well justified in the topical report.
Specific language in the conclusions and how that language was supported by information in the topical report was another subject of discussion. Specifically, the precision of the phrases a range of normal operating and off-normal accident conditions, satisfactory performance, and fission product release data and fuel failure fractions, as summarized in this report, can be used for licensing of reactors and the context of what is being requested to be approved were discussed. NRC staff pointed out that although these phrases could be used in the report, they could result in limitations or conditions being imposed by the NRC in the safety evaluation of the report, while a more precise request for these terms could avoid those potential limitations.
Based on discussions with the applicant, further detail in these areas were deemed to be more suitable for requests for additional information (RAIs). Applicant and industry participants expressed a desire to have follow-up discussions on these areas before RAIs are issued.
NRC staff also toured INL facilities where post irritation examination of the fuel was performed at the Materials and Fuel Complex. Applicant and industry participants showed staff hot cells, fuel holding areas, examination and sorting equipment and assessment methods for examining tested fuel down to the level of an individual TRISO particle. While these methods are not the subject of the requests in the topical report, the context and visuals were instrumental in providing staff with context and better understanding the data in the report.
Finally, the applicant identified a transcription error (value in the table inconsistent with the source) in the SiC density in Table 5-5, to be updated as part of a future submittal. NRC staff had extensive interactions with various applicant and industry participants during the course of the audit and practically all of the questions asked were answered in some fashion. On the whole, the submitted topical report was backed by well-established documentation. The areas identified above represent topics where additional clarity could be established by the applicant.
H.
Exit Briefing The NRC staffs audit exit meeting was conducted on October 9, 2019. The vendor and NRC staff discussed the staffs takeaways from the audit, and the staff indicated the necessary information within the scope of the audit was provided to assist the staff in their review.
I.
Requests for Additional Information (RAIs) Resulting from Audit As part of the staff review, the staff expects there to be RAIs in the following areas:
- 1. justification that the performance parameters referenced in Conclusion 1 sufficiently bound the data set,
- 2. justification that the physical parameter envelope in Table 5-5 and discussed in Conclusion 2 is sufficient to rely on to provide satisfactory fuel performance,
- 3. limitations of the topical report to fuel manufactured using an uninterrupted coating process,
- 4. basis for the specified ranges of acceptable fuel in Table 5-5 that are beyond the tested ranges,
- 5. specificity related to the statement in Conclusion 3 that, fission product release data and fuel failure fractions, as summarized in this report, can be used for licensing of reactors employing UCO TRISO-coated fuel particles that satisfy the parameter envelope defined by measured particle layer properties in Table 5-5, and
- 6. description to capture the adequacy of the QA program used to collect the data supporting the topical report.
Participants from the applicant and industry expressed a desire to have a public meeting on the areas identified by the staff before RAIs are issued.
J.
Open Items and Proposed Closure Paths No open items were identified as a result of the audit.
K.
Deviations from the Audit Plan No deviations from the audit plan were identified or required.