ML26046A093
| ML26046A093 | |
| Person / Time | |
|---|---|
| Site: | 05000614 |
| Issue date: | 02/13/2026 |
| From: | NRC |
| To: | NRC/NRR/DANU |
| References | |
| +reviewed | |
| Download: ML26046A093 (0) | |
Text
{{#Wiki_filter:From: ProjectLongMott-Safety Sent: Friday, February 13, 2026 8:23 PM To: ProjectLongMott-SafPUBLICem Resource
Subject:
[External_Sender] TRISO-X Pebble Fuel Qualification Methodology LTR Attachments: 2023-XE-NRC-018-A PUBLIC.pdf; [External_Sender] TRISO-X Pebble Fuel Qualification Methodology LTR_Redacted.pdf
Hearing Identifier: XeDOW_ProjectLongMott_SafPublic Email Number: 261 Mail Envelope Properties (SA1PR09MB96928ADB1C2DEC3285B7A11ABD6EA)
Subject:
[External_Sender] TRISO-X Pebble Fuel Qualification Methodology LTR Sent Date: 2/13/2026 8:23:10 PM Received Date: 2/13/2026 8:23:14 PM From: ProjectLongMott-Safety Created By: ProjectLongMott-Safety.Resource@nrc.gov Recipients: "ProjectLongMott-SafPUBLICem Resource" <ProjectLongMott-SafPUBLICem.Resource@nrc.gov> Tracking Status: None Post Office: SA1PR09MB9692.namprd09.prod.outlook.com Files Size Date & Time MESSAGE 3 2/13/2026 8:23:14 PM 2023-XE-NRC-018-A PUBLIC.pdf 7252513 [External_Sender] TRISO-X Pebble Fuel Qualification Methodology LTR_Redacted.pdf 127148 Options Priority: Normal Return Notification: No Reply Requested: No Sensitivity: Normal Expiration Date:
X Energy, LLC 801 Thompson Avenue Rockville, MD 20852 +1 301.358.5600 www.x-energy.com Page 1 of 2 9 June 2023 2023-XE-NRC-018 Project No. 99902071 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Submittal of X Energy, LLC (X-energy) Topical Report, "TRISO-X Pebble Fuel Qualification Methodology" Accepted Version
REFERENCES:
- 1) Letter from M. van Staden to U.S. Nuclear Regulatory Commission (NRC) dated 30 April 2021, Submission of X Energy, LLC (X-energy) Xe-100 Topical Report: TRISO-X Pebble Fuel Qualification Methodology (ML21120A334)
- 2) Letter from Travis Chapman to NRC dated 2 September 2021, Re-Submission of X Energy, LLC (X-energy) Xe-100 Topical Report: TRISO-X Pebble Fuel Qualification Methodology (ML21246A289)
- 3) Email to X Energy LLC, Transmittal of Requests for Additional Information - Xe-100 Topical Report:
TRISO-X Pebble Fuel Qualification Methodology (ML21334A230)
- 4) Letter from Travis Chapman to NRC dated 28 February 2022, Submission of X Energy, LLC (X-energy)
Responses to Requests for Additional Information for Topical Report, "TRISO-X Pebble Fuel Qualification Methodology" (ML22059A770)
- 5) Letter from Travis Chapman to NRC dated 29 July 2022, Submission of X Energy, LLC (X-energy) TRISO-X Pebble Fuel Qualification Methodology (ML22216A179)
- 5) Letter from NRC to T. Chapman dated 9 March 2023, X ENERGY, LLC - FINAL SAFETY EVALUATION FOR XE-100 TOPICAL REPORT: TRISO-X PEBBLE FUEL QUALIFICATION METHODOLOGY (REVISION 3) (EPID NO.
L-2021-TOP-0011/CAC NO. 000431) (ML22327A198) In References 1 and 2 listed above, X Energy, LLC (X-energy) submitted Topical Report TRISO-X Pebble Fuel Qualification Methodology for NRC review. In Reference 3, NRC transmitted Requestsfor Additional Information (RAIs), which X-energy responded to in Reference 4. In Reference 5, X-energy submitted Revision 3 of the subject report. The purpose of this letter is to submit the accepted version of the subject topical report. This report contains commercially sensitive, proprietary information and, as such, we are requesting that this information be withheld from public disclosure in accordance with 10 CFR 2.390, "Public inspections, exemptions, request for withholding," paragraph (a)(4). Additionally, certain information in this report was determined to contain Export Controlled Information (ECI). This information must be protected from disclosure pursuant to 10 CFR 810. Enclosure 1 is the Non-Public version of the report with all attachments which contains nonredacted sensitive, proprietary information that is appropriately marked. Enclosure 2
X Energy, LLC 801 Thompson Avenue Rockville, MD 20852 +1 301.358.5600 www.x-energy.com Page 2 of 2 provides an affidavit with the basis for this request. Enclosure 3 provides a redacted copy of the report with all attachments that contains non-proprietary content. This letter contains no commitments. If you have any questions or require additional information, please contact Ingrid Nordby at inordby@x-energy.com. Sincerely, Travis A. Chapman Director, Reactor Licensing and Regulatory Affairs X Energy, LLC cc: X-energy, LLC David Bannister Steve Miller Martin van Staden Marcy Sanderson U.S. Nuclear Regulatory Commission Andrew Proffitt Stephanie Devlin-Gill Michael Orenak U.S. Department of Energy Jeff Ciocco Carl Friesen
Enclosures:
- 1) Xe-100 Topical Report, TRISO-X Pebble Fuel Qualification Methodology, Accepted Version (Proprietary)
- 2) Affidavit Supporting Request for Withholding from Public Disclosure (10 CFR 2.390)
- 3) Xe-100 Topical Report, TRISO-X Pebble Fuel Qualification Methodology, Accepted Version (Non-Proprietary)
X Energy, LLC 801 Thompson Avenue Rockville, MD 20852 +1 301.358.5600 www.x-energy.com X Energy, LLC Xe-100 Topical Report, TRISO-X Pebble Fuel Qualification Methodology Accepted Version (Proprietary)
X Energy, LLC 801 Thompson Avenue Rockville, MD 20852 +1 301.358.5600 www.x-energy.com Affidavit Supporting Request for Withholding from Public Disclosure (10 CFR 2.390)
X Energy, LLC 801 Thompson Avenue Rockville, MD 20852 +1 301.358.5600 www.x-energy.com Page 1 of 2 Affidavit Supporting Request for Withholding from Public Disclosure (10 CFR 2.390) I, Travis A. Chapman, Director, Reactor Licensing & Regulatory Affairs, Xe-100 Program, of X Energy, LLC (X-energy) do hereby affirm and state:
- 1. I am authorized to execute this affidavit on behalf of X-energy. I am further authorized to review information submitted to or discussed with the Nuclear Regulatory Commission (NRC) and apply for the withholding of information from disclosure. The purpose of this affidavit is to provide the information required by 10 CFR 2.390(b) in support of X-energys request for proprietary treatment of certain commercial information submitted in Enclosure 1 to X-energys letter XE-NRC-2023-018 from myself to the NRC which provides the accepted version of the X-energy Topical Report, TRISO-X Pebble Fuel Qualification Methodology for the X-energy Xe-100 Nuclear Reactor.
- 2. I have knowledge of the criteria used by X-energy in designating information as sensitive, proprietary, confidential, and export-controlled.
- 3. Pursuant to the provision of paragraph (b)(4) of 10 CFR 2.390, the following is furnished for consideration by the NRC in determining whether the information sought to be withheld from public disclosure should be withheld.
- a. The information sought to be withheld from public disclosure in Enclosure 1 is owned by X-energy. This information was prepared with the explicit understanding that the information itself would be treated as proprietary and confidential and has been held in confidence by X-energy.
- b. The information sought to be protected in Enclosure 1 is not available to the public.
- c. The information contained in Enclosure 1 is of the type that is customarily held in confidence by X-energy, and there is a rational basis for doing so. The information X-energy is requesting to be withheld from public disclosure includes technical information related to the design, analysis and operations associated with our Xe-100 high-temperature, gas-cooled, pebble bed advanced reactor design that directly impact our business development and commercialization efforts. X-energy limits access to this proprietary and confidential information in order to maintain confidentiality.
- d. Enclosure 1 contains information about the planned activities of X-energy related to the development of the Xe-100 design bases, TRISO-X fuel design bases, forecast design development timeframes, and relate to the commercialization strategy for our Xe-100 advanced reactor. Public disclosure of the information contained in Enclosure 1 would create substantial harm to X-energy because it would reveal valuable technical information regarding X-energys design development, competitive expectations, assumptions, current position and strategy. Its use by a competitor could substantially improve the competitors position in the design, manufacture, licensing, construction and operation of a similar competing product.
X Energy, LLC 801 Thompson Avenue Rockville, MD 20852 +1 301.358.5600 www.x-energy.com Page 2 of 2
- e. Additionally, Enclosure 1 is assessed to contain certain information that is considered Export Controlled Information (ECI) under the provisions of 10 CFR 810. I have personal knowledge of the criteria used by X-energy to evaluate documents for ECI and affirm that this information should be withheld from public disclosure.
- f. The Proprietary Information contained in Enclosure 1 is transmitted to the NRC in confidence and under the provisions of 10 CFR 2.390; it is to be received in confidence by the NRC. The information is properly marked.
I declare under the penalty of perjury that the foregoing is true and correct. Executed on July 9, 2023. Sincerely, Travis Chapman Director, Reactor Licensing & Regulatory Affairs, Xe-100 Program X Energy, LLC
X Energy, LLC 801 Thompson Avenue Rockville, MD 20852 +1 301.358.5600 www.x-energy.com X Energy, LLC Xe-100 Topical Report, TRISO-X Pebble Fuel Qualification Methodology Accepted Version (Non-Proprietary)
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Z/KEh>/ ^/'EZ/d Z/ y &h >WZKhd^W /&/d/KE 5 COATED-PARTICLE FUEL EXPERIENCE BASE..................................................................................... 53
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6 XE-100 FUEL QUALIFICATION PROGRAM...................................................................................... 118
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7 NRC TOPICAL REPORT REVIEW OBJECTIVES.................................................................................. 135 8 REFERENCES................................................................................................................................. 136 APPENDIX A. NRC ASSESSMENT OF NGNP WHITE PAPERS: FOLLOW UP ITEMS FOR FUEL QUALIFICATION........................................................................................................................... 144 APPENDIX B. STATISTICAL QC METHODS......................................................................................... 159 APPENDIX C. FUEL SPHERE TESTING................................................................................................ 163 APPENDIX D. REQUESTS FOR ADDITIONAL INFORMATION & RESPONSES........................................ 164
WŽ y dŽZŽ dZ/^KyW&YŽDŽŽŽ ŽEŽ ZŽ : Ž y >> ŽŽŽy ZZ??Z??> WŽ W Ž >Ž>dZ FIGURES & y^Žs & dZ/^K&WŽŽŽ&Ž & yZ& & W,d'ZZŽZŽ^ Figure 5: Logic for Deriving Fuel Product Specification & ZŽ & ŽW & d/ŽŽdZ/^KŽW & W & KWŽdŽsŽK & dZ/^KŽW&D & <DŽ, K& & >Ž&ŽWŽŽ^ & KŽŽŽŽ^Ž> hhK dZ/^K& & dŽŽŽŽ^Ž & W & W & W & ZŽ<Ž'> hhK dZ/^K& & ZŽ<Ž'> hhK dZ/^K& & &ŽDZŽsZ^> hhK & & D<ZŽ,dZWDWŽŽd&,&ZW & h^W& & <ZŽZ,dŽ,d'Z&d & WŽ ŽŽYŽŽ'Z& & ,ŽŽŽŽŽ'Z & 'ZW'&ŽW & <ZŽ'ZŽŽ,Žh^&Z' &'ZŽ^Ž'Z & Z&WŽ'Zd,Ž & ^Ž'ZEŽŽŽ & Z&WŽ'Zd,Ž
WŽ y dŽZŽ dZ/^KyW&YŽDŽŽŽ ŽEŽ ZŽ : Ž y >> ŽŽŽy ZZ??Z??> WŽ W Ž >Ž>dZ & &ŽWŽDŽ'Z & ŽsŽ'Z/WDŽ & &ŽWŽZŽ'ZŽ & ZŽ'ZŽ & ŽŽŽhKhK WŽ'Z^d & ^&ŽŽW/ WŽŽŽ & W & W & W &WŽŽZŽŽŽD<
WŽ y dŽZŽ dZ/^KyW&YŽDŽŽŽ ŽEŽ ZŽ : Ž y >> ŽŽŽy ZZ??Z??> WŽ W Ž >Ž>dZ TABLES d^Wy&^Ž d^^dWŽŽYhK^ dyZŽD d<dŽ>ZŽŽŽZŽy dWy&Z dZŽ&&D&WWŽ dEŽKŽWŽŽ^/Ž^ d^Ž,d<ŽZZ d Ž/Ž&ŽŽ/Ž,d dWWDZ&YŽWŽ d'Z&WŽ/Žd dYDŽŽhKdZ/^K&Ž dhK<Ž'Z& dWŽŽdZ/^KŽhKWŽ'Z& d^WŽŽ'ZŽ d'Zd&/ŽŽŽ d'Z/Žd^Ž d <KZW/,dŽ> hhK d ŽŽŽ^& d W d W d W d dŽŽŽ'Z/ŽdŽhKŽ dEZE'EW&ŽŽh/ dE^^ŽWŽd d ZŽŽ/ŽŽZŽ
WŽ y dŽZŽ dZ/^KyW&YŽDŽŽŽ ŽEŽ ZŽ : Ž y >> ŽŽŽy ZZ??Z??> WŽ W Ž >Ž>dZ ABBREVIATIONS dŽŽŽ Abbreviation or Acronym Definition Z^ ŽŽŽZŽ^ hE 'Z 'ZŽ E^ E^Ž KK KŽK E^/ EŽ^/ Z ZŽ ^D ^ŽŽD dZ dZŽ/E> sZ sŽ & ŽŽŽŽ Ž /^K /ŽŽ &Z ŽŽ&ZŽ K> Ž> s ŽŽŽ E E >K& >ŽŽ&ŽŽŽ K Ž d& dŽ& ŽŽ ^ ŽŽŽ^ &W Ž W ŽWŽŽ WZ/ WŽZ/ W? W?Ž Ž
WŽ y dŽZŽ dZ/^KyW&YŽDŽŽŽ ŽEŽ ZŽ : Ž y >> ŽŽŽy ZZ??Z??> WŽ W Ž >Ž>dZ Abbreviation or Acronym Definition &/D &Ž/DŽ &K< Ž &Y &YŽ &^s &Ž^s ' 'Ž '^^Z 'Ž^^ZŽ ' ' 'dD,Z 'dDŽ,ZŽ 'tDdh 'DdŽh 'h/ 'h/ , ,ŽŽ ,&Z ,&ZŽ ,D ,D ,Dd
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WŽ y dŽZŽ dZ/^KyW&YŽDŽŽŽ ŽEŽ ZŽ : Ž y >> ŽŽŽy ZZ??Z??> WŽ W Ž >Ž>dZ Abbreviation or Acronym Definition Dd^ ŽŽ Dt D Dt D E'EW E'ŽEW EWZ EWŽŽZŽ EY EY EZ EZŽŽŽ KW KWŽŽ KZE> KZEŽ>ŽŽ W' WŽŽ' WDZ WDŽZŽ W/ ŽŽŽ W/, ŽŽ W/Zd WŽ/ŽZd WZ WŽZ W^/ W^/ŽŽŽ Z ŽŽ Z/ ZŽŽ/ŽŽ Z^ ZŽŽŽ^ Z' ZŽ' ZE Ž ZWs ZŽWs ^&> ^Z ^ZŽ ^ZZ> Ž ^D,Z ^DŽ,ZŽ ^ Ž ^ZD ^ZDŽ ^ZW ^ZW d ŽŽ d dŽ Ž
WŽ y dŽZŽ dZ/^KyW&YŽDŽŽŽ ŽEŽ ZŽ : Ž y >> ŽŽŽy ZZ??Z??> WŽ W Ž >Ž>dZ Abbreviation or Acronym Definition d,dZ dŽ,dZŽ dZ/^K d/ŽŽ hK ŽŽh hK ss ŽŽ s,dZ s,dZŽ y y >>
WŽ y dŽZŽ dZ/^KyW&YŽDŽŽŽ ŽEŽ ZŽ : Ž y >> ŽŽŽy ZZ??Z??> WŽ W Ž >Ž>dZ
- 1. INTRODUCTION y ŽŽŽy Ž,d'ŽŽZŽ,d'Z
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ŽŽŽ ŽŽ x
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WŽ y dŽZŽ dZ/^KyW&YŽDŽŽŽ ŽEŽ ZŽ : Ž y >> ŽŽŽy ZZ??Z??> WŽ W Ž >Ž>dZ 1.3. INTERFACING REFERENCES dŽŽ ŽŽŽŽŽ ŽŽ EZŽy Ž y Ž Ž Ž Ž D^d ŽŽŽŽy ŽŽŽ ŽŽŽŽŽŽ ŽE / ŽEZZŽ'Z' EZ dŽŽŽ Ž x WZ/ŽŽŽdZ/^KŽŽ'ZŽ WZ/Z EZ^ Ž^ ŽŽ WZ/Z&^ Z x D ^Ž d &Ž Ž Ž y ŽŽy ZZ??Z??> >Ž Ž x WŽ Z WZ Ž Ž Ž Ž^D E^Z^WŽZ^ŽEŽ> tZŽEWŽWZ^ x > > Ž Ž^^ŽŽ^^ Ž Ž Ž ŽŽŽ E / yŽŽy ZZ??Z??> Z/ŽWŽ>ŽŽy ZŽs ZŽŽydŽŽŽdZŽ ŽŽy ŽŽEZŽŽ Ž 1.4. DOCUMENT LAYOUT ^Ž ŽŽŽŽŽy Ž ^Ž ŽŽŽŽŽdZ/^KŽŽŽ,d'Z ^Ž Žy ^Ž ŽŽŽŽdZ/^KŽŽ Ž Ž Ž dZ/^K ŽŽ Ž h^ h <Ž 'Z:^Ž^Ž Ž ŽŽy ^Ž ŽEZŽŽŽ dŽŽŽŽŽŽ ŽEZŽE'EW&YŽD^Žd ŽŽŽYŽŽYŽŽŽŽ dZ/^KŽŽ ŽŽ W/ŽŽ ŽŽ ŽŽ
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- 2. BACKGROUND 2.1. XE-100 REACTOR DESIGN d,d'ZŽŽŽŽŽŽŽKŽŽŽd
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Žy h^/EZ ŽŽŽ>tZ ŽŽŽ,d'Z ŽŽŽŽŽŽ Ž >tZ,d'ZŽŽZEZK Ž ŽŽŽŽ Ž>tZŽŽŽŽ ŽŽŽŽŽŽ ŽŽ ŽdŽŽŽ Ž y Ž Ž ŽdŽŽŽ &ZŽ&Z ŽŽ 3.1. NRC REGULATIONS EZŽŽŽŽŽŽŽ Ž Ž Ž /ŽŽŽŽ ŽŽ ZŽ Ž Ž Ž Ž ŽŽ ŽŽŽKKŽ''Ž Ž &Z W d ' Ž ŽŽ Ž Ž ŽŽŽŽ d' ŽŽŽ >tZŽŽŽŽŽ Ž,d'Z EŽŽŽŽ&ZW 'ŽŽŽŽŽŽ ŽŽŽ ZŽŽ'Reactor Design 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. K Ž Ž ŽŽ Ž Ž ^&>Ž'ŽŽŽŽKK ''ŽŽŽŽ >tZŽŽŽ ^ŽŽ ŽŽŽŽŽ ZŽ Ž'ŽŽŽŽ>tZ EZ K ŽŽ Ž Ž Ž Ž Ž ŽKŽŽŽŽ ŽŽ ZŽZŽŽŽŽ>tZdŽŽ ŽŽŽ,d'Z WŽ y dŽZŽ dZ/^KyW&YŽDŽŽŽ ŽEŽ ZŽ : Ž y >> ŽŽŽy ZZ??Z??> WŽ W Ž >Ž>dZ K ŽŽŽŽŽZŽEZ/E> EZ ŽŽŽEZ ZŽ'Ž ŽŽ>tZ EZŽŽEZ dŽŽŽZŽ' 'ŽŽWŽEŽ>tZŽ d Ž Ž ŽŽŽ,d'Z dŽ&ZWŽŽŽ ŽŽŽŽdŽŽ y Ž,d'ZŽŽZ ,d'Z Ž'ŽŽ Reactor design. The reactor system and associated heat removal, control, and protection systems shall be designed with appropriate margin to ensure that specified acceptable system radionuclide release design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences. dŽŽŽŽŽŽŽŽ Ž Design features within the reactor system must ensure that the SARRDLS are not exceeded during normal operations and AOOs. The (TRISO) fuel used in the MHTGR design is the primary fission product barrier and is expected to have a very low incremental fission product release during AOOs. As noted in NUREG-1338 and in the NRC staffs feedback on the Next Generation Nuclear Plant (NGNP) project white paper, Next-Generation Nuclear Plant - Assessment of Key Licensing Issues the TRISO fuel fission product transport and retention behavior under all expected operating conditions is the key to meeting dose limits, as a different approach to defense in depth is employed in an MHTGR. The SARRDL concept allows for some small increase in circulating radionuclide inventory during an AOO. To ensure the SARRDL is not violated during an AOO, a normal operation radionuclide inventory limit must also be established (i.e., appropriate margin). The radionuclide activity circulating within the helium coolant boundary is continuously monitored such that the normal operation limits and SARRDLs are not exceeded. The SARRDLs will be established so that the most limiting license-basis event does not exceed the siting regulatory dose limits criteria at the exclusion area boundary (EAB) and low-population zone (LPZ), and also so that the 10 CFR 20.1301 annualized dose limits to the public are not exceeded at the EAB for normal operation and AOOs. ,d'ZŽŽWŽŽŽŽŽ ŽŽdZ/^KŽ Containment design. A reactor functional containment, consisting of multiple barriers internal and/or external to the reactor and its cooling system, shall be provided to control the release of radioactivity to the environment and to ensure that the functional containment design conditions important to safety are not exceeded for as long as postulated accident conditions require. dŽŽŽŽŽŽŽŽŽŽ WŽ y dŽZŽ dZ/^KyW&YŽDŽŽŽ ŽEŽ ZŽ : Ž y >> ŽŽŽy ZZ??Z??> WŽ W Ž >Ž>dZ The term functional containment is applicable to advanced non-LWRs without a pressure retaining containment structure. A functional containment can be defined as a barrier, or set of barriers taken together, that effectively limit the physical transport and release of radionuclides to the environment across a full range of normal operating conditions, AOOs, and accident conditions. Functional containment is relied upon to ensure that dose at the site boundary as a consequence of postulated accidents meets regulatory limits. Traditional containment structures also provide the reactor and SSCs important to safety inside the containment structure protection against accidents related to external hazards (e.g., turbine missiles, flooding, aircraft). The MHTGR functional containment safety design objective is to meet 10 CFR 50.34, 52.79, 52.137, or 52.157 offsite dose requirements at the plants exclusion area boundary (EAB) with margins. The NRC staff has brought the issue of functional containment to the Commission, and the Commission has found it generally acceptable, as indicated in the staff requirements memoranda (SRM) to SECY-93-092 and SECY-03-0047. In the SRM to SECY-03-0047 (Ref. 10), the Commission instructed the staff to develop performance requirements and criteria working closely with industry experts (e.g., designers, EPRI, etc.) and other stakeholders regarding options in this area, taking into account such features as core, fuel, and cooling systems design, and directed the staff to submit options and recommendations to the Commission for a policy decision. The NRC staff also provided feedback to the DOE on this issue as part of the NGNP project. In the NRC staffs Next Generation Nuclear Plant Assessment of Key Licensing Issues, the area on functional containment and fuel development and qualification noted that approval of the proposed approach to functional containment for the MHTGR concept, with its emphasis on passive safety features and radionuclide retention within the fuel over a broad spectrum of off-normal conditions, would necessitate that the required fuel particle performance capabilities be demonstrated with a high degree of certainty. yŽŽŽŽŽŽ ŽŽ /Ž Ž &Ž Ž ZŽ y E'??'>?? t ŽŽŽŽŽŽ Žd Ž Ž Ž Ž Ž ŽŽŽŽŽŽ dŽŽŽŽ E^ŽŽ ŽŽŽ dŽŽŽ ŽŽŽŽŽŽŽ^^y Ž KŽŽ &ZŽ ŽŽŽŽŽ>tZ&ZŽŽ ŽŽŽŽ ŽŽŽ ^d ^ŽŽŽŽŽŽ ŽŽŽŽŽŽŽŽ ŽŽŽ ŽŽŽŽŽŽŽ ,ŽŽ,d'Zy ŽŽŽŽŽ>tZ ^ W Ž Ž Ž Ž Ž WŽ y dŽZŽ dZ/^KyW&YŽDŽŽŽ ŽEŽ ZŽ : Ž y >> ŽŽŽy ZZ??Z??> WŽ W Ž >Ž>dZ ŽŽŽŽ ŽŽŽŽ> &Z &ZŽŽŽŽŽK> Ž Ž Ž Ž Ž Ž Ž Ž>tZŽŽŽŽ ŽŽŽŽ &Z ŽŽ Ž Ž Ž Ž ŽŽŽŽŽŽ dŽŽy ŽdZ/^KŽ Ž ŽŽŽ 3.2. NRC POLICY STATEMENTS EŽEZŽŽŽŽŽŽy Ž ŽŽŽŽŽŽEZŽŽŽŽ ŽEZ Ž,ŽEZŽ ŽŽ,d'Z^ zEZ ^Ž ŽEhZ ' EZŽŽŽŽ ŽŽ Ž Ž^ z Ž^ZDKŽ^ zŽ ŽWŽ^ŽdŽŽ Ž,d'ZŽŽŽ ŽŽ ŽŽŽŽ Ž 3.2.1. Functional Containment Performance Criteria &ŽŽ,d'ZŽZEŽ ŽŽŽ ŽŽŽŽd>tZŽ ':Ž&ZW dy Ž Ž ŽŽŽŽŽdZ/^K ŽŽŽ ŽŽ /^ZDŽ^ZDŽ^ zEZŽŽ EZŽŽŽŽŽŽŽ ŽŽ Ž/^ zEZ ^ zEZ ŽŽ Ž Ž Ž Ž Ž ŽŽŽ ŽŽŽŽ ŽŽ ^ŽŽŽ ŽE'EWŽŽŽEZŽE'EW ŽŽŽ,d'ZŽŽŽŽŽEZ /^ zEZ EZŽŽŽŽŽŽŽŽ ŽŽŽ^^ ŽŽ Ž Ž Ž WŽ y dŽZŽ dZ/^KyW&YŽDŽŽŽ ŽEŽ ZŽ : Ž y >> ŽŽŽy ZZ??Z??> WŽ W Ž >Ž>dZ ŽE /dŽŽŽŽŽŽ Ž Ž ^^ Ž Ž Ž Ž Ž Ž Ž ŽŽZEdEZŽŽŽŽŽŽŽ Ž^ZD ^ zŽŽŽŽŽŽ ŽŽŽŽŽŽ,d'ZŽŽdŽŽŽ Ž Ž E / Ž Z' d ŽŽŽŽ> Žy ŽŽ EŽ>t ZŽ E>tZ WZ Ž Ž Ž Ž Ž Ž Ž Ž ŽŽ> dŽŽŽ ŽŽŽ&dE /ŽŽŽŽŽŽŽ Ž ŽŽŽ> Ž /y ŽŽ> Ž^^ŽŽ^^ ^&>Ž^ZZ> ^^ ŽŽ ŽŽŽŽ Ž E /y ŽŽŽŽs dŽŽŽŽ ŽŽŽ ŽŽŽŽ^^y ŽŽ Ž ŽŽ x ^ŽŽŽŽŽŽ x 'ŽŽŽ x WŽŽ x W x ^ŽŽ 3.2.2. Source Term dŽŽŽŽ,d'Z Ž ŽZEŽŽŽŽ Ž> d,d'ZŽŽŽŽŽŽ ŽŽ ŽŽŽŽ /^ z ŽŽŽŽŽŽŽŽ ŽŽŽŽŽŽŽŽŽ Ž ŽŽ ŽŽ ŽŽ Ž dŽŽŽŽŽŽ ŽŽŽŽŽŽ dŽD^d Ž^ zEZE>tZWZŽ ^ŽŽD E^Ž^D E^Ž Z WŽ y dŽZŽ dZ/^KyW&YŽDŽŽŽ ŽEŽ ZŽ : Ž y >> ŽŽŽy ZZ??Z??> WŽ W Ž >Ž>dZ ^E / ŽEZdŽŽ yD^dŽŽ>Ž 3.3. NRC GUIDANCE/REFERENCES 3.3.1. NUREG-0111, Evaluation of High Temperature Gas-Cooled Reactor Particle Coating Failure Models and Data EhZ 'dŽ ŽŽŽ , hh dZ/^K dK ŽŽ/^K Ž,d'Z Ž Ž ' Ž DŽ Ž ŽŽŽŽŽŽy Ž Ž Ž Ž > h d Ž Ž EhZ 'ŽŽŽŽŽ EŽ Ž Ž Ž Ž,d'Z ŽŽŽŽŽ> hhK dZ/^KŽy Ž 3.3.2 NUREG-0800, Standard Review Plan (LWR Edition), Section 4.2, Fuel System Design dŽŽEhZ '^ZW^ZW ^ŽEZ Ž>tZŽ' ŽŽŽŽ&ZdŽ^ZWŽŽ ^&> >tZŽŽŽŽŽ ŽŽŽ ŽŽŽŽŽŽ ŽŽŽŽŽ ŽŽd^&>^ZW^ŽŽ>tZ hK Ž ^Ž ŽŽŽŽŽ,d'ZŽŽ ŽEZŽŽZŽŽ> Ž E / dŽŽ^&>^ZWEZ x Ž x ŽŽ x ŽŽ x WŽŽ ŽŽŽŽ^ZW ŽŽŽ>tZŽŽŽ WŽ y dŽZŽ dZ/^KyW&YŽDŽŽŽ ŽEŽ ZŽ : Ž y >> ŽŽŽy ZZ??Z??> WŽ W Ž >Ž>dZ ŽŽŽŽŽŽŽŽ ŽŽŽ,d'Z ^ZWŽŽŽ ,ŽŽ^ZWŽ>tZ/EhZ 'W EZŽEhZ 'ŽŽŽŽŽ Ž,Ž^ŽEhZ 'ŽŽ ŽŽŽ ^ZWŽŽŽŽ ŽŽdŽŽŽ ŽŽŽŽŽd ŽŽŽŽŽŽtŽŽŽ ŽŽŽŽŽŽ &ŽyŽŽŽ Ž ŽŽ Ž ZE Ž ŽŽŽ /ŽŽŽ / ^ d WŽ Ž Ž y ŽŽŽŽŽŽŽŽ ŽEZ E'EWWŽŽŽ^ŽKZEŽ >ŽŽKZE>ŽŽEZŽŽŽ^Ž Ž Ž^Z^^Z^ŽE'EWŽ,d'Zd^ZW ŽŽŽŽ^Z^dKZE>Ž ŽE'EWWŽdKZE>ŽŽ Ž ŽŽŽŽ ŽŽŽŽŽ 3.3.3 TRISO-Coated Particle Fuel Phenomenon Identification and Ranking Tables / Ž Ž Ž Ž,d'Z EZ ŽŽ Ž Ž Ž dZ/^KŽ Ž Ž ŽŽdZ/^KŽŽ ŽŽŽŽŽŽEhZ 'ZsŽW/Zd ^WŽ/ŽZdW/ZdŽ W/ZdŽ x D x KŽ x x Z x x Ž WŽ y dŽZŽ dZ/^KyW&YŽDŽŽŽ ŽEŽ ZŽ : Ž y >> ŽŽŽy ZZ??Z??> WŽ W Ž >Ž>dZ /W/ZdŽŽhK Ž W/ZdŽŽ,ŽŽ ŽŽŽŽ ŽhK ŽŽEhZ 'ZEZŽ W/ZdŽ /ŽŽŽŽŽŽ WŽŽŽŽ,d'ZŽ WŽŽŽŽ Ž Ž Ž Ž ŽŽ ŽŽ /ŽŽŽŽEZ,d'ZŽŽ ŽŽŽŽ ^ŽŽŽEZŽŽŽŽ WŽŽŽŽ,d'Z yŽŽŽŽW/ZdŽ Ž Ž Ž Ž dZ/^K ŽŽŽŽ'ZWŽ 3.4 ADDITIONAL GUIDANCE: ANS-53.1 dE^ŽE^^E^^WŽŽ DŽ,ŽŽZŽWE^Ž dŽŽEZŽŽŽŽ ŽŽ ŽE / EŽŽ ŽŽŽ,d'ZŽŽŽ ŽŽŽŽEZŽŽŽ ŽŽŽŽŽ E^ŽŽ ŽŽŽŽŽŽŽ ŽŽŽdy ŽŽ Ž Ž 3.5 U.S. HTGR PRECEDENTS 3.5.1 Peach Bottom 1 ŽŽŽW ŽŽWŽŽh,d'Z ŽŽŽ:dDtŽ Ž Ž , hŽ /^KŽ Ž Ž WŽ y dŽZŽ dZ/^KyW&YŽDŽŽŽ ŽEŽ ZŽ : Ž y >> ŽŽŽy ZZ??Z??> WŽ W Ž >Ž>dZ ŽŽdŽŽ Ž Ž Ž Ž DŽŽŽŽdZ/^K 3.5.2 Fort St. Vrain d&Ž^s&^sE'^Ž,d'Z Dt ŽŽŽDt&^sŽŽŽ>ŽŽ&^s Ž dZ/^K Ž Ž , h Ž 3.5.3 GASSAR 'Ž^^ZŽ'^^Z'Ž Ž Ž ŽŽ,d'Z Dt Dt '^^Z d Ž ŽŽW ŽŽ&Ž^Ž/Ž EZŽddZ/^K'^^Z, h dZ/^KŽŽ/^KEZŽ ŽŽŽ'^^Zd&ŽŽ ŽŽŽEZŽŽZ^d ŽŽŽŽŽ,d'ZŽ ŽŽŽ&W&Z/Ž &ŽEZŽŽ&WŽŽdŽ 3.5.4 MHTGR Conceptual Design Preliminary Safety Information Document Review /EZ EhZ 'ŽŽŽW^ ZEZ ŽEZŽŽ'ŽDŽ, d 'ŽŽ ZŽ D,d'Z ŽŽ Ž d ŽD,d'ZdZ/^KŽ> hhKŽ Žy&ŽŽŽŽEhZ 'K ŽŽŽ ŽŽEZŽŽ ŽŽŽW^ ZŽEZ dŽW^ ZŽW^ ZŽŽŽ ŽW^ Z dW^ ZŽŽŽŽŽŽŽW^ Z Ž x d EZ Ž Ž Ž ŽŽŽK ŽŽŽ ŽŽŽŽŽ WŽ y dŽZŽ dZ/^KyW&YŽDŽŽŽ ŽEŽ ZŽ : Ž y >> ŽŽŽy ZZ??Z??> WŽ W Ž >Ž>dZ x dEZŽŽ&ZŽ'&Z'Ž ŽŽŽŽŽŽŽ Ž ,ŽEhZ 'Ž ŽŽŽŽŽ,d'ZŽ ŽŽŽŽŽŽ/ ŽŽŽŽEZ ŽŽ Ž x ŽŽŽŽŽ ŽŽŽŽŽ x YŽŽŽŽŽŽŽŽ x &ŽŽŽŽŽŽŽŽ x ŽŽŽŽ Ž x WŽŽŽŽŽŽŽ Ž EhZ 'ŽŽŽŽŽŽŽ y x dŽŽŽŽŽ ŽŽŽŽ x d Ž ŽŽ Ž D,d'Z EZ Ž Ž ŽD,d'Z>Ž ŽEZŽŽŽŽ ŽŽŽ x d Ž Ž Ž Ž ŽŽŽŽŽŽŽ ŽŽŽŽŽ Ž EhZ 'Ž ŽŽZŽŽŽŽ ŽEhZ 'ŽŽŽ ZŽŽŽŽ x dŽŽŽŽŽŽŽ Ž x dŽ Ž ŽŽŽŽŽ x sŽŽŽŽŽŽ WŽ y dŽZŽ dZ/^KyW&YŽDŽŽŽ ŽEŽ ZŽ : Ž y >> ŽŽŽy ZZ??Z??> WŽ W Ž >Ž>dZ d W^ ZŽŽ ŽŽ'Z&ŽYŽWŽ/E> ŽŽ^ŽDŽŽŽŽ W^ ZE'EW ŽŽŽ^Ž 3.5.5 Pebble Bed HTGRs dŽŽEZŽ,d'Z Ž Ž ŽŽ / Ž 'Ž Ž Ž ŽEZŽŽŽ5 ŽŽŽ&ZdŽ > hhK dZ/^KŽEZ ŽŽ^ zEZ Ž^ z ŽŽŽŽ Ž ŽŽ ŽWŽEZŽ ŽŽ^/ ŽEZ ŽŽŽdEZŽŽ ŽŽŽ Ž ŽŽŽ ŽŽEZ ,ŽŽŽŽEZŽŽŽŽ Ž ŽEZ ŽŽ Ž'> hhK ŽŽ / ^Ž Ž WDZ W > Ž Ž ŽŽEZŽŽŽŽŽ,Ž ŽWDZW>Ž 3.5.6 Next Generation Nuclear Plant (NGNP) Preapplication Review /h^K E'EWWŽ/ŽEŽ>ŽŽ/E>dŽ ŽŽŽŽŽŽ,d'ZŽŽ E'EWŽŽŽ,d'ZŽŽŽŽŽ Ž d Ž Ž Ž K EZ Ž Ž Ž K EZ Ž ŽŽ Ž Ž K EZ d Ž Ž ŽŽŽŽŽ ŽE'EWŽŽŽŽ/E> 5 /ŽWDZŽŽ^ŽŽŽWŽŽŽ ŽŽŽ WŽ y dŽZŽ dZ/^KyW&YŽDŽŽŽ ŽEŽ ZŽ : Ž y >> ŽŽŽy ZZ??Z??> WŽ W Ž >Ž>dZ dE'EWWŽ,d'ZEZŽ ŽE'EWŽŽ/ŽŽŽŽ Ž ŽŽŽŽŽ Ž,d'ZŽŽEZ d Ž ŽE'EWŽZEŽ ŽdZ/^KŽŽŽŽ Ž Ž Ž d Ž ,d'Z Ž EZZŽWŽ^ŽŽŽŽ ŽŽŽŽŽŽ ŽŽŽŽŽŽŽŽŽ ZE Ž E'EW Ž Ž ŽŽ Ž Ž ŽŽŽŽŽŽ E'EWŽŽ K E'EW>tŽ'Žd E'EW>tŽ'ŽŽŽ,d'Z ŽŽŽŽŽŽ/E>E'EWZŽŽ ŽEZŽŽŽŽ&YŽD ^Žd> ^ŽdŽ Ž Ž ŽŽ Ž Ž Ž Ž EZŽŽŽŽ E'EWŽEZŽŽŽ &ŽŽŽŽŽŽ&YŽD^Ž dtŽ> ^Ž^Ž Ž ^ ^ ŽŽ EZ / EZ ŽŽE'EWE'EWK /E>EZŽ Ž Ž Ž ŽŽ Ž Ž Ž Ž Ž d Ž ŽŽŽŽŽEZW ŽŽ ŽŽŽŽ,d'ZdŽŽ &ŽŽŽ >Ž ^Ž &ŽŽŽŽŽŽŽ /DEZŽŽŽ ŽŽŽWdŽŽ EZŽŽŽZŽ^Z^Z^ DZ^ŽŽŽEZZ^Ž ŽŽŽŽEZ/ŽŽŽ Z^ŽŽŽŽŽŽ Ž WŽ y dŽZŽ dZ/^KyW&YŽDŽŽŽ ŽEŽ ZŽ : Ž y >> ŽŽŽy ZZ??Z??> WŽ W Ž >Ž>dZ dEZŽŽŽ &YŽD^Žd:EZ Ž> ^Ž^ŽŽ ^^ŽŽŽ ŽŽ ŽŽŽ Ž E / dEZŽE'EWŽŽŽŽŽ ,d'ZŽŽŽŽdŽŽŽZE ŽŽŽŽŽŽ ŽŽŽŽ Ž Ž Ž ZE ŽŽ d Ž ŽŽ^ zŽ^ZD dŽŽŽEZŽ /E>'ZŽ'Z&WŽEZŽ ŽŽ ŽŽŽŽ Ž ŽŽŽŽŽdZ/^KŽ ŽŽ Ž Ž,d'Z Ž Ž Ž Ž Ž Ž Ž E'EW Ž Ž KŽ'Z&WŽŽEZŽŽE'EW Ž Ž Ž Ž Ž Ž ŽdZ/^KŽd^Ž WŽ y dŽZŽ dZ/^KyW&YŽDŽŽŽ ŽEŽ ZŽ : Ž y >> ŽŽŽy ZZ??Z??> WŽ W Ž >Ž>dZ 4 FUEL DESIGN AND PERFORMANCE REQUIREMENTS dŽŽ Žy dŽy Ž ŽŽ ^Ž D ^Ž ŽŽŽŽŽŽŽ ŽŽŽŽŽ Ž ŽdŽŽŽŽ^Ž ŽŽŽŽŽŽŽŽ ŽŽŽ 4.1. 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WŽ y dŽZŽ dZ/^KyW&YŽDŽŽŽ ŽEŽ ZŽ : Ž y >> ŽŽŽy ZZ??Z??> WŽ W Ž >Ž>dZ Figure 28: Kr-85m R/B Data for AGR-1/-2 Compared to Historical U.S. and FRG Data dZŽ'ZŽ& 'ZŽŽ hKŽŽŽŽYŽŽ ŽhK ŽW/ ŽŽŽŽhKŽŽ ŽŽŽŽŽ ŽŽŽŽ ŽdŽŽŽŽŽ hK dZ/^KŽ Ž Ž Ž Ž Ž Ž Ž Ž Ž Ž Ž,Ž d'ZŽhKdZ/^KŽŽŽŽ Ž Ž Ž,d'Z ŽŽ Ž,d'Z,Ž d Ž 'Z Ž ŽŽ Ž Ž Ž Ž Ž Ž,d'Z Ž d Ž ŽŽ'ZŽŽ ŽŽ 5.3.2.2.2 AGR-3/4 Test dŽŽ'ZŽŽ'ZŽ'ZŽ/ Ž Ž &W Ž Ž dZ/^KŽ ŽŽ ŽŽŽ,d'ZŽdŽ ,d'ZŽŽŽdZ/^KŽŽd 'ZŽŽŽdZ/^KŽ d&dhKŽŽŽŽ 'Zdd&ŽŽ ŽŽŽŽŽŽŽ Ž WŽ y dŽZŽ dZ/^KyW&YŽDŽŽŽ ŽEŽ ZŽ : Ž y >> ŽŽŽy ZZ??Z??> WŽ W Ž >Ž>dZ d'ZŽŽŽŽ'ZŽ ŽŽŽŽŽŽŽŽŽŽ Ž Ž & Ž Ž Ž Ž ŽŽŽŽŽhKŽ d&hK Figure 29: AGR-3/4 Cross-Section (left) and an AGR-3/4 Capsule (right) ŽŽ'ZŽŽŽŽhK ŽZŽŽŽ ŽŽ,d'Z15 'ZŽŽŽ ŽŽŽ'Z/Ž ŽŽŽŽŽŽdŽ Ž Z Ž 'Z 'Z Ž Ž Ž Ž ŽŽ ŽŽ&W& ŽZŽ 'Z'ZŽ dZ ŽŽŽŽŽŽŽŽ ,ŽZ Ž 15dŽ&WŽŽ'ZŽŽŽŽŽhK ŽŽŽ,ŽŽ ŽW/ ŽŽŽ WŽ y dŽZŽ dZ/^KyW&YŽDŽŽŽ ŽEŽ ZŽ : Ž y >> ŽŽŽy ZZ??Z??> WŽ W Ž >Ž>dZ Figure 30: R/B per Failed Particle Data from AGR 2 and 3/4 Tests and Historical Data 5.3.2.2.3 AGR-5/6/7 Test d 'Z W Ž Ž Ž 'Z Ž'ZŽŽhKdZ/^KŽŽ'Z ŽD D hŽ'ZŽ Ž'Z Ž & ŽŽŽŽ ŽŽ Ž,d'ZŽŽŽŽdZ/^KŽ&Ž'ZŽ Ž ŽŽ ŽŽ ŽŽ Ž ŽŽ&Ž'Z Ž ŽŽ'Z ŽdZŽ ŽŽŽŽŽŽd ŽŽŽ &/DŽ&/D ŽŽ Ž &W Figure 31: Schematic view of the AGR-5/6/7 test train (Note: Capsule 5 is at the top of the test train). WŽ y dŽZŽ dZ/^KyW&YŽDŽŽŽ ŽEŽ ZŽ : Ž y >> ŽŽŽy ZZ??Z??> WŽ W Ž >Ž>dZ dŽŽŽ&Ž Ž/E>dZh ŽŽŽŽ ŽŽ ŽŽŽŽ:ŽŽ ŽŽŽŽ ŽŽ 'ZŽŽŽ &W ŽdZ Z ŽŽ Ž ŽŽŽŽ Table 15. AGR-5/6/7 Irradiation Test Specifications & ŽZŽ'ZŽ WŽ y dŽZŽ dZ/^KyW&YŽDŽŽŽ ŽEŽ ZŽ : Ž y >> ŽŽŽy ZZ??Z??> WŽ W Ž >Ž>dZ Figure 32: R/B per Failed Particle Data from AGR 2, 3/4, & 5/6/7 Tests and Historical Data 5.3.2.3 PIE and Safety Testing dŽŽW/ Ž ŽŽdZ/^KŽ ŽŽŽ dŽŽŽŽ ŽŽŽŽŽŽ ŽW/ ŽŽŽŽŽŽ Ž Ž Ž Ž Ž Ž Ž ŽŽ W/ ŽhKdZ/^K'ZŽ ŽŽ'ZdW/ ŽŽŽŽ ŽŽ <ŽŽ &WŽŽŽ ŽŽ ^ŽŽ ŽŽŽŽŽ Ž WŽ y dŽZŽ dZ/^KyW&YŽDŽŽŽ ŽEŽ ZŽ : Ž y >> ŽŽŽy ZZ??Z??> WŽ W Ž >Ž>dZ 5.3.2.3.1 Fission Product Distributions Ž Ž Ž Ž &W Ž Ž ŽŽŽŽ'ZŽŽ ŽŽŽŽŽ ŽŽŽŽ^Žd & Ž Ž Ž Ž Ž ŽŽŽ ŽŽŽ ŽsŽŽŽŽŽ ŽŽŽŽŽ^Ž &W Figure 33: Fission Product Mass Balance for AGR-1 Capsules & ŽŽ /ŽŽ ^ŽŽ ŽŽ^ŽŽŽ ZŽ^ ŽŽ ŽŽ ŽŽŽ ŽŽŽŽ Ž,ŽŽŽŽŽŽ ŽŽŽŽ ŽŽŽŽ WŽ y dŽZŽ dZ/^KyW&YŽDŽŽŽ ŽEŽ ZŽ : Ž y >> ŽŽŽy ZZ??Z??> WŽ W Ž >Ž>dZ dŽWŽŽ^ŽŽ ŽŽWŽWŽŽŽŽŽ^Ž Ž Ž Ž Ž dZ/^KŽ d WŽ^dZ/^KŽŽ ŽD ^ŽŽŽŽŽ ŽŽ 5.3.2.3.2 Irradiated Fuel Particle Microstructural Evaluation ŽŽŽŽŽŽŽdŽ ŽŽ Ž Ž WŽ ŽŽ ŽŽ Ž Ž Ž Ž Ž / Ž Ž Ž Ž ŽŽŽŽŽŽŽŽ ŽŽŽ Ž& d ŽŽŽhKŽ ŽŽŽŽŽŽ /ŽŽ'ZŽŽ/WdŽ Ž& Ž& ŽŽD ŽŽ/WŽŽ Ž& Ž Ž Ž ^ Ž Ž Ž ŽŽŽŽŽ &ŽŽŽŽŽŽŽŽ ŽŽŽŽŽŽŽŽ Ž& tŽŽ& ŽŽŽŽŽŽŽŽŽ Ž/WKŽ /W& ŽŽ Ž/WŽŽŽŽ/WŽ^& &ŽŽŽ/WŽŽ/Ž ŽŽ/WŽŽŽ/WŽ & ŽŽ/W^ Ž^/W^Ž& Ž Ž ^ /W Ž & Ž ŽŽ Ž Ž& Ž/WŽŽŽŽ& ^Ž^, Ž /W Ž & Ž & d ŽŽŽŽŽŽ^Ž/W^Ž ^ Ž WŽ y dŽZŽ dZ/^KyW&YŽDŽŽŽ ŽEŽ ZŽ : Ž y >> ŽŽŽy ZZ??Z??> WŽ W Ž >Ž>dZ ŽŽŽŽŽŽŽŽŽ /WŽŽŽŽ ^ŽŽ/WŽŽ /WŽŽW/ Ž'ZŽŽŽ /ŽŽŽ/WŽ 'Z ŽŽŽŽ/WŽŽ /Ž ŽŽ'ZŽŽ ŽŽ Figure 34: Examples of Various AGR-1 Irradiated Particle Microstructures EŽŽŽŽŽŽŽ ŽŽŽŽŽŽ'ZW/ ŽŽŽ&/D ŽŽŽ ŽŽŽ^ 5.3.2.3.3 Safety Testing WŽŽŽŽhKdZ/^KŽ Ž'ZŽŽŽhKdZ/^KŽ&WŽ Ž& sŽ WŽ y dŽZŽ dZ/^KyW&YŽDŽŽŽ ŽEŽ ZŽ : Ž y >> ŽŽŽy ZZ??Z??> WŽ W Ž >Ž>dZ ŽŽDŽ ^Ž Ž'ZŽŽ& ŽŽ Ž Ž ^ d ŽŽŽ^ ŽŽŽ Figure 35: Fission Product Release from AGR-1 Compact 6-4-3 at 1600°C. 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KZE>dDKZEŽ>ŽŽ X Energy, LLC dŽŽ ZŽD x-energyŽ W9Ž11 &ŽŽŽŽ ŽŽWŽ Ž Ž Ž & y Ž Ž ŽŽ ŽŽŽŽ ŽŽŽŽŽŽ y Ž ŽŽŽŽ W Ž Ž ŽŽŽŽŽŽ X Energy, LLC dŽŽ ZŽD x-energyŽ W10Ž11 NRC Review ofdZ/^KyW&YŽDŽŽŽZŽ NRC RAI Question Number dŽ&ZŽWŽ ŽŽdZŽŽ,d'Z Žy dh^EZŽŽŽEZ dZy ŽWŽ,d'Z Ž ŽŽŽŽŽŽŽ ŽŽŽ ŽŽŽŽŽŽŽ ŽŽŽŽ WŽŽŽŽ/dZy EZŽdŽ^Ž ŽŽŽŽyŽ ŽdZŽŽEZŽŽ ŽŽŽ WŽŽ&ŽŽ ŽŽŽ ^y ŽŽŽŽ(( ))P, ŽŽŽŽ/ ŽŽŽŽŽŽŽ ŽŽ ŽŽŽŽ ŽŽŽŽ'Z ŽŽŽŽŽŽ ŽŽŽ (( ))P X-energy Response &Ž&YŽDŽŽŽZŽŽŽŽydZ/^K WŽ/Ž ŽW/ WŽŽŽ X Energy, LLC dŽŽ ZŽD x-energyŽ W11Ž11 yŽŽŽŽ(( ))PŽ &Ž d'ZŽŽyd'ZŽŽŽ hKdZ/^KŽŽŽ dŽŽŽŽŽŽŽ Ž'ZŽŽŽDW ŽŽDW/ŽŽŽŽ ŽŽŽdZ/^KŽ'Z ŽŽ(( ))P, E,ŽŽ ŽŽŽŽŽ ŽdZ/^K(( ))P (( ))P, E Z'ZyŽŽW/ ŽŽ ŽŽ'ZŽŽŽ ŽŽy (( ))P OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Travis Chapman Director, Reactor Licensing and Regulatory Affairs X Energy, LLC 801 Thompson Avenue Rockville, MD 20852
SUBJECT:
X ENERGY, LLC - FINAL SAFETY EVALUATION FOR XE-100 TOPICAL REPORT: TRISO-X PEBBLE FUEL QUALIFICATION METHODOLOGY (REVISION 3) (EPID NO. L-2021-TOP-0011/CAC NO. 000431)
Dear Mr. Chapman:
By letter dated April 30, 2021, X Energy, LLC (X-energy) submitted topical report (TR) Xe-100 Topical Report: TRISO-X Pebble Fuel Qualification Methodology, Revision 1, (Agencywide Document Access and Management System (ADAMS) Accession No. ML21120A332) for the U.S. Nuclear Regulatory Commission (NRC) staffs review. This TR summarizes the fuel qualification approach and methodology for the tri-structural isotropic (TRISO)-coated particle fuel pebbles to be used in the Xe-100 high temperature gas-cooled reactor design under development by X-energy. A closed meeting was held on June 4, 2021, to provide the NRC staff with an overview of the TR (ML21354A289). On September 2, 2021 (ML21246A289), X-energy submitted Revision 2 of the TR for the NRC staffs review, which was updated from Revision 1 based on the NRC staffs feedback provided during the June 4, 2021, closed meeting. The NRC staff reviewed the TR and issued requests for additional information (RAIs) on December 2, 2022 (ML21334A230). A closed meeting was held on February 10, 2022 (ML22166A418), to clarify the RAIs. The NRC staff received RAI responses on February 28, 2022 (ML22059A770), and a closed meeting was held on May 20, 2022 (ML22166A424), to discuss the RAI responses. By letter dated July 29, 2022 (ML22216A179), X-energy submitted Revision 3 of the TR for the NRC staffs review, which addressed the RAIs and RAI responses related to Revision 2 of the TR. By letter dated September 21, 2022 (ML22265A218), X-energy requested the withdrawal of Revisions 1 and 2 of the TR. Accordingly, the NRC staff continued its review of Revision 3 of the TR. On November 7, 2022, the NRC staff provided X-energy a draft of the safety evaluation (SE) to ensure the correct marking of proprietary information (ML22194A948). X-energy confirmed the marking of proprietary information in a letter dated November 30, 2022 (ML22363A285). The final NRC staff SE for Revision 3 of the TR is enclosed. A redacted version of the SE will be made publicly available. In accordance with the NRCs published information for TRs available at https://www.nrc.gov/about-nrc/regulatory/licensing/topical-reports.html, the NRC staff requests March 9, 2023
OFFICIAL USE ONLY - PROPRIETARY INFORMATION T. Chapman OFFICIAL USE ONLY - PROPRIETARY INFORMATION that X-energy publish an accepted version of this TR within 3 months of the receipt of this letter. The accepted version should incorporate this letter and the enclosed SE. Also, the accepted version must contain historical review information, including the RAIs and RAI responses. The accepted version should include an -A (designated accepted) following the TR identification number. As an alternative to including the RAIs and RAI responses, if changes to the TR were provided to the NRC staff to support the resolution of RAI responses, and the NRC staff reviewed and approved those changes, there are two ways that the accepted version can capture the RAIs: 1. The RAIs and RAI responses can be included as an Appendix to the accepted version. 2. The RAIs and RAI responses can be captured in the form of a table (inserted after the final SE) which summarizes the changes as shown in the accepted version of the TR. The table should reference the specific RAIs and RAI responses which resulted in any changes, as shown in the accepted version of the TR. If you have any questions or comments concerning this matter, please contact Michael Orenak at (301) 415-3229 or via e-mail at or Michael.Orenak@nrc.gov. Sincerely, William Jessup, Chief Advanced Reactor Licensing Branch 1 Division of Advanced Reactors and Non-Power Production and Utilization Facilities Office of Nuclear Reactor Regulation Project No.: 99902071
Enclosure:
As stated cc via ListServe: Distribution at X-Energy Xe-100 The Enclosure to this letter contains Proprietary Information. When separated from the Enclosure, this letter is DECONTROLLED. Signed by Jessup, William on 03/09/23
Package: ML22327A199 Letter: ML22327A198
Enclosure:
ML22327A202 NRR-106 OFFICE NRR/DANU/UAL1:PM NRR/DANU/UAL1:PM NRR/DANU/UAL2:LA OGC NAME KWagner MOrenak CSmith JEzell DATE 02/06/2023 02/16/2023 03/08/2023 09/30/2022 OFFICE NRR/DANU/URT1:BC NRR/DANU/UAL1:BC NAME MHayes WJessup DATE 10/17/2022 3/9/2023
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION X ENERGY, LLC - SAFETY EVALUATION OF XE-100 TOPICAL REPORT, TRISO-X PEBBLE FUEL QUALIFICATION METHODOLOGY, REVISION 3 (EPID NO. L-2021-TOP-0011) SPONSOR AND SUBMITTAL INFORMATION Sponsor: X Energy, LLC. (X-energy) Sponsor Address: X Energy, LLC. 801 Thompson Ave Rockville, MD 20852 Project No.: 99902071 Submittal Date: July 29, 2022 Submittal Agencywide Documents Access and Management System (ADAMS) Accession No.: ML22216A179 RAI response letter date and ADAMS Accession No: February 28, 2022 (ML22059A770) Brief Description of the Topical Report: Xenergy is developing a high-temperature gas-cooled reactor (HTGR) called the Xe100, which uses TRI-structural ISOtropic (TRISO)-coated Uranium Oxycarbide (UCO) fuel in pebble form. One key aspect of the safety case for the design relates to the role of the TRISO fuel and its associated functional requirements. The topical report (TR) provides information intended to ultimately lead to the qualification of TRISO-coated particle fuel pebbles for use in the Xe-100 reactor. As described in the TR, X-energy states that the fuel qualification program is intended to cover the fuel specifications, implemented fabrication processes and associated quality controls, and demonstrated fuel performance under heated and irradiated conditions. The TR provides context related to the role of the TRISO fuel in the Xe100 design, identifies regulations and guidance that X-energy expects to be applicable to the TRISO fuel as used in the Xe-100 design, and provides historical context on data and experience on TRISO fuel. This background is then used to explain the activities that X-energy plans to perform to qualify the TRISO fuel for the Xe100, including the top-level design and performance requirements. Ultimately, the TR provides a methodology and plan for qualifying fuel for the Xe100 design. The TR requests that the NRC staff find that fuel design and performance requirements in Section 4 [of the TR] are adequate for establishing an acceptable design basis to support the licensing of the Xe-100 reactor and plans established in Section 6 [of the TR] for qualification of the UCO TRISO-coated particles in spheres are generally acceptable. March 9, 2023
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION While this safety evaluation (SE) documents the NRC staffs review of Revision 3 of the TR, this SE refers to RAI responses in multiple locations. For RAI responses that did not result in updates to the TR but nonetheless contained information that was used in this SEs findings, X-energy included the responses as an enclosure to Revision 3 of the TR. REGULATORY EVALUATION Regulatory Basis For construction permit applications, Title 10 of the Code of Federal Regulations (10 CFR) paragraphs (a)(1)(ii)(D)(1) and (a)(1)(ii)(D)(2) of Section 50.34, Contents of applications; technical information, apply to the Xe100 design because the fuel is the primary means of fission product retention. Similar regulatory requirements exist for design certification applications, combined license applications, and standard design approvals (10 CFR 52.47(a)(2)(iv)(A) and (B), 10 CFR 52.79(a)(1)(vi)(A) and (B), and 10 CFR 52.137(a)(2)(iv)(A) and (B), respectively). Also, 10 CFR 50.34(a)(3)(i) requires, in part, that an applicant for a construction permit to build a power reactor will provide principal design criteria (PDC) for the facility. These PDC should be informed by the general design criteria (GDC) in 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, Appendix A, General Design Criteria for Nuclear Power Plants, as the GDC are considered to be generally applicable to other types of nuclear power units and are intended to provide guidance in establishing the principal design criteria. Although the GDC applies only to light-water reactor (LWR) designs, the NRC staff expects that non-LWR designs will have PDC that fulfill a similar role. Similar regulatory requirements exist for design certification applications, combined license applications, 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). The PDC establishes requirements for structures systems and components (SSCs) that are important to safety. GDC 10, Reactor design, in Appendix A to 10 CFR Part 50 states 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. Examples of PDC for various advanced reactor designs can be found in Regulatory Guide (RG) 1.232, Guidance for Developing Principal Design Criteria for Non-Light-Water Reactors, (ML17325A611) which provides guidance for developing PDC for non-LWR designs. Establishing fuel design limits and ensuring that these limits are not exceeded, represent a fundamental underpinning of the safety assessment of a nuclear power plant required by 10 CFR 50.34(a)(1). Other GDC relate to fuel design and radionuclide limits during plant operation. The NRC staff anticipates that the TR will be used, in part, to fulfill requirements related to PDC developed by the applicant in support of the limits imposed by similar design criteria. It is the NRC staffs understanding that the TR will form a portion of the basis for meeting these regulations. The NRC staff reviewed the TR, as described below, against these regulations in conjunction with the overall technical acceptability of the TR within the requested scope of the TR.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION TECHNICAL EVALUATION In the following NRC staff evaluation, a section, table, or figure number without an additional description refers to the TR. Additional descriptive information is provided if section, table, or figure numbers are referenced from other reports or sections of this SE. Methodology Overview The TR focuses on a subset of TRISO fuel qualifications for the Xe100 design. TRISO fuel in the Xe100 reactor is in pebble form, with dimensions and relevant operating parameters identified in Table 1. Xenergys design and safety philosophy relies partly on these parameters falling within the envelope of the previously reviewed and approved Electric Power Research Institute (EPRI) TRISO TR and associated SE (TR References 2 and 7, respectively). These philosophies are reflected in the functional containment approach proposed by X-energy for the Xe100 design. A functional containment can be defined as a set of barriers taken together that effectively limit the physical transport and release of radionuclides to the environment across a full range of operating and accident conditions. The TRISO particles form a key barrier to the release of radionuclides that form a substantial portion of the functional containment approach, as shown in Figure 4. Fuel qualifications for the TRISO particles, in this case, involve demonstrating that fuel particles maintain a substantial (here, specified as 99.999 percent) proportion of fission products within the particle under reactor operating conditions. In effect, the NRC staff understands that the goal of the TR is to provide a linkage demonstrating that the proposed fuel testing gives assurance that the asmanufactured fuel will meet the fuel performance goals provided in the TR under operating and accident conditions. The methodology for qualifying fuel to specified standards involves identifying asmanufactured fuel quality requirements (effectively, a manufacturing envelope within which a substantial defined fraction of particles must fall, as well as a statistical means for evaluating whether the population of particles meets the specification based on samples from the manufacturing process) as well as inservice fuel performance requirements (a range of operating and/or accident conditions that the fuel must remain within to provide assurance of effective operation). Figure 5 provides an illustration of the iterative process that X-energy used to develop the manufactured fuel specification, and Table 3 provides preliminary full particle fuel performance release criteria for both defective and intact particles. The NRC staff notes that individual particle parameters are not directly provided, and this is discussed further below in the Utilization of the Advanced Gas Reactor (AGR) UCO Compact Data section of this SE. The parameters derived from Table 3 are then used as requirements for asmanufactured quality and inservice performance of coated-particle fuel. The TR states that these parameters are then iterated on as shown in Figure 6. The qualification of fuel is not limited to just the TRISO particles, however. In addition to the diffusion releases of some fission products discussed later in this SE, fission products can also be released as a result of contamination of the fuel pebbles during manufacturing and some fission product release can occur from volatile elements that are not entirely failed. The overall, comprehensive fuel qualification plan is summarized in Section 6, but the TR makes up only a part of that comprehensive fuel qualification. This SE identifies areas where the TR does not fully address the entirety of the fuel qualification process via limitations and makes no findings
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION on areas outside the scope of the TR (some of which are identified below in the Portions of the TR Not Addressed by this SE section of this SE). Fuel Design and Performance Requirements As described in the TR, the methodology is intended to demonstrate that the asmanufactured fuel (yet to be produced) will perform under operating and accident conditions in accordance with values derived from testing, which serve as inputs to calculated radionuclide release values that serve as design and acceptance criteria. At a high level, this process is shown in Figure 6. Final values for sub-steps in this process are not identified (and are inferred to be beyond the scope of the TR). Preliminary fuel requirements related to quality, expected inservice failure fractions, and allowable releases are provided in Section 4. The TR states that Production of high-quality TRISO-coated particle fuel is achieved through a combination of product specifications, QC methods, and manufacturing equipment specifications, each of which plays an important role. These and other specifications are generally identified as draft in the TR. These specifications yield a calculated fission product release value from intact particles, based on the information in Table 3 and Section 4.2. Manufacturing defects and contamination from non-graphite sources are another source of fission product release and these are addressed in the TR. Table 17 provides a mean expected value for contamination and silicon carbide (SiC) defects that would contribute to the source term. The test program discussed in the TR and below in the Role of Confirmatory Testing section of this SE would provide indications of particle performance under operating and accident conditions, but this would only be of use for the particles themselves (as well as potentially the defect values for the tested particles). Based on the TR, the NRC staff understands that the values in Table 17 for contamination and SiC defects would effectively serve as inputs to the source term calculation and would be based on asmanufactured measured parameters. In addition, other physical parameters not directly related to thermal or nuclear performance (such as mechanical effects) are stated, as part of the response to RAI 4, to be outside the scope of the TR. Use of the methodology outlined in the TR therefore relies on a justification relating a measured manufactured value to a corresponding acceptable source value to be used in a future licensing application. In part, this represents Limitation 1 on the TR. Section 7 of the TR defines fuel qualification as: Establishment of fuel product, equipment, and feedstock specifications; Implementation of a fuel fabrication process capable of consistently and reliably meeting the specifications at the required scale; Implementation of statistical [quality control/quality assurance] procedures to demonstrate that the product specifications have been met; Irradiation of statistically sufficient quantities of fuel with monitoring of inpile performance and post-irradiation examination to demonstrate that normal operation performance requirements are met; and
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Post-irradiation heating tests (safety testing) of statistically sufficient quantities of irradiated fuel to demonstrate that accident condition performance requirements are met. The NRC staff considers this definition, together with the supporting documentation provided in the TR, to represent a fuel qualification methodology. This methodology, which combines particle performance with other influences on fission product releases to derive an effective fuel performance, is a sound approach that captures the relevant influences on radionuclide release from TRISO-based fuel. Apart from the final two bullets (discussed further in the next paragraph), the NRC staff finds that the methodology is appropriate to develop a thermal and nuclear design basis for the Xenergy TRISO fuel, but by itself, is not sufficient to qualify fuel, nor does it represent a design basis, which would be provided as part of a licensing basis for a full plant design or facility license application. Demonstration of the final two bullets, as applicable to the Xenergy limiting reactor parameters (yet to be finalized), represents a necessary step in qualifying the specific fuel to be used in the Xenergy design. However, because the testing is only planned at this stage and has not been executed, the TR is only an intermediate step in qualifying the fuel. The testing is discussed further below in the Role of Confirmatory Testing section of this SE. Therefore, the NRC staff finds that this approach provides an acceptable foundational framework for qualifying fuel, with the final demonstration of fuel qualification requiring outstanding items to be provided as part of future design submittals (e.g., a final fuel design specification, the operating performance envelope of the Xe100 and how the provided fuel fits within that envelope). Other efforts, such as those identified in response to RAI 4, are necessary to demonstrate the performance under accident conditions (e.g., transient behavior). Further, as discussed above, parameters not related to thermal or nuclear parameters (such as mechanical effects) are stated to be outside the scope of the TR. These parameters have traditionally been considered within the scope of fuel qualification but are not identified in the list above. As such, the scope of the TR approval is limited to the plan to test TRISO fuel and the operational envelope outlined in the TR. Other aspects associated with fuel qualification, including how the final Xe100 fuel meets the criteria outlined in the TR should be included as part of future licensing submittals that reference the TR for the Xe100 design. This represents Limitation 2 on the TR. The NRC staff notes that the methodology highlights the role of uncertainties in both particle parameters (explicit as part of the statistical methodology for evaluating whether particles meet their requirements) and reactor operational values (which would be more fully captured as part of future submittals on the design and would identify how the fuel design falls within the operational envelope). Adequately assessing the uncertainty is important in ensuring that there is no gap between theoretical fuel performance and asbuilt fuel performance. The TR does not precisely identify the relationship between the manufacturing specification and the reactor design envelope. Therefore, a finding on the adequacy of this subject was judged to be outside the scope of the TR. Utilization of the AGR UCO Compact Data Section 5 provides an overview of the experience associated with coated TRISO particles. It covers the historical fabrication process as it has evolved over time and provides a detailed discussion of the failure mechanisms of particles, with some discussion on probable causes of
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION those failure mechanisms. However, the TR does not seek to directly tie those failure mechanisms to specific reactor parameters - rather, failures are aggregated and grouped based on the operational envelope which the particles are subject to, with temperature being a primary parameter of interest. This analytical approach, in effect, affords an applicant the capability to calculate a failure fraction based on empirical results rather than directly modeling the particles. Subsequently, an applicant can use this approach to either directly determine radionuclide releases from the statistical population of particles subject to various reactor conditions or to validate analytical tools based on these empirical results. The NRC staff finds this approach to be acceptable for developing a baseline fuel qualification output based on the testing already performed and proposed as part of the TR. Further refinement of what is included in this analytical approach is anticipated to be the subject of future licensing submittals. The final fuel form for the Xe100 is TRISO particles in pebbles. As the TR notes, there is extensive experience historically with both compact-and pebble-based TRISO fuel, but the AGR program (which the TR references) testing was performed using cylindrical compacts. In order to confirm that the Xenergy process for manufacturing pebbles performs similar to the AGR fuel form, X-energy states that it plans to test pebble fuel in a form with specifications corresponding to those planned for the Xe100 design. Although the TR provides top-level radionuclide requirements for the fuel in Table 2, and aggregate full particle specification requirements in Table 3, no specifications for what constitutes acceptable ranges for the particle parameters in the Xe100 are provided (though AGR2 values are provided in Tables 11 and 12). The TR references, but does not explicitly incorporate, the previously reviewed and approved EPRI TRISO TR and associated SE (TR References 2 and 7 respectively). The TR states [w]hile the Xe-100 design has evaluated several fuel specifications that could be used safely in operation, the reference Xe-100 fuel particle chosen is the same as the 15.5%-enriched, 425-Om UCO-coated particle irradiated in the AGR-5/6/7 qualification/margin test. In its response to RAI 1, Xenergy states the complete determination [of fuel quality and performance requirements] and underlining methodologies are beyond the scope of this report and will be addressed in future topical reports. With respect to the design performance envelope, the TR provides Figure 25, which falls within the performance envelope outlined in the aforementioned EPRI TRISO TR. These values are preliminary in this case. The TR also provides UCO kernel attributes and properties for AGR2 fuel in Tables 11 13. Because the TR is not clear on the fuel performance envelope for the final design and the particle specifications planned for use in the Xe100 are not clearly defined in the TR, the NRC staff is imposing the limitations and conditions associated with the EPRI TRISO TR (as documented in TR Reference 7) as Limitation 3. In short, these relate to: the qualification of the fuel form as opposed to the particles themselves; the content and makeup of the TRISO particle as compared to the AGR program tested particles, as these tests are referenced by the EPRI TRISO TR; the performance envelope associated with the AGR test particles and; the disposition of short-lived fission products as part of the source term.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION The NRC staff expects that the testing planned to be performed as discussed in the TR will resolve these limitations and conditions (with the possible exception of short-lived fission products, which should be addressed as part of a future licensing submittal). However, it is the responsibility of an applicant referencing the TR, to justify how the particle specification for a future application submittal either falls within an approved performance envelope (e.g., the reviewed and approved TRISO TR) or can be shown to be acceptable (e.g., by drawing parallels between known acceptable particles and tested particles that match the application fuel specification). Further discussion of the planned testing is provided below. In general, the qualification approach in the TR relies on the AGR test program with a specific focus on the parameters and results contained in Tables 11, 12, and 13, and Figures 25, 30 and
- 33. The approach used in the TR conforms with the conclusions in the SE for the EPRI TRISO TR (TR Reference 7). In particular, the NRC staff agrees with the statements in Section 5.3.2.4, while noting that [n]o full TRISO-coated particle failures have been observed at 1600°C or 1700°C does not suggest that there are no releases from particles at those temperatures, as evidenced from the AGR testing and discussed in the TR and the EPRI TRISO TR (Reference 2). The NRC staff finds that the AGR testing forms a baseline that, when augmented by the confirmatory testing discussed below and in the TR, provides that an applicant referencing the TR with an empirical basis for the UCO TRISO fuel is to be used in subsequent design and licensing submittals.
Role of Confirmatory Testing The qualification of TRISO fuel has generally been performed based on two linked types of tests: direct irradiation testing in a reactor and post-irradiation testing of fuel at higher temperatures in different environments to simulate post-accident conditions (sometimes referred to as safety testing). The AGR program testing that is referenced in the TR utilizes this approach. X-energy plans to provide further demonstration of adequate fuel performance through its own testing, as described in the TR. While extensive testing has been performed on TRISO fuel in relatively inert environments (generally helium, as in the AGR tests), testing data for TRISO-based fuel subject to other environments that may present challenges during accident conditions is more limited. As a result, in order to provide additional assurance of an acceptable performance under a broader range of potential accident conditions, X-energy plans to perform additional post-irradiation safety testing (( )) Section 5.1.5.5 and Appendices B and C, discuss quality controls on the fuel form to be tested. The NRC staff understands that these quality controls will provide assurance that the tested fuel pebbles and particles will meet the acceptance criteria associated with intact fuel elements. The NRC staff finds that the use of the quality control and sampling means discussed in the TR, is acceptable. This acceptance is limited to the parameters of the fuel to be produced for testing. The NRC staff notes that these measures do not represent a commitment for the final Xe100 design. Future licensing submittals referencing the TR should provide a set of specifications and demonstrate how fuel to be used in the reactor will meet these specifications. This is reflected in Limitation 2.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION The TR requests the NRC staffs review and approval of the statement [p]lans established in Section 6 for qualification of the UCO TRISO-coated particles in spheres are generally acceptable. In the response to RAI question 3, X-energy states that (( ] and that tests to be used by the Xe100 code evaluations and models are based on the data that shows extremely limited failures at burnups of up to approximately 20 percent fissions per initial metal atom at 1800 degrees Celsius (°C) (3272 degrees Fahrenheit (°F)). The NRC staff has reviewed the TR using the information provided in the RAI responses in conjunction with the information in the TR to make the findings below. (( ] Conditions for these planned tests are documented in Tables 19 and 20. As part of the response to RAI question 5, X-energy stated that the test process would involve following the procedures laid out in Figure 38. These procedures, described further in the TR, are consistent with the procedures used to gather data to qualify TRISO fuel (as was performed in the AGR program). (( )) The NRC staff finds that the results of the tests described in Tables 19, 20 and 21 can be used as part of the demonstration of fuel performance under the test conditions for the final design, subject to the limitations identified below. Because neither the final design parameters (including accident environments and temperatures) nor the results of the tests are available, a future licensing application referencing the TR, leveraging these test results, will need to provide a justification for how the reactor accident conditions are bounded by the test results (which, due to test uncertainties, may not precisely match the conditions outlined in the tables, hence the need to provide a justification). In referencing previously performed testing, the TR states that [t]he overall consistency of these data should adequately address the concern raised by the NRC staff about potential effects of neutron spectral differences between water-cooled [Materials Testing Reactors] and an actual operating modular HTGR. This conclusion appears reasonable, but the NRC staff makes no
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION finding with respect to testing aside from the applicability of the AGR program testing and the testing planned to be performed as discussed in the TR. Based on the discussion in the TR, the NRC staff finds the proposed testing to be applicable to the currently described iteration of the Xe100, and the results can be used as part of qualifying the Xe100 fuel. The NRC staff makes no findings related to the results of those tests at this time because they are not yet performed and available. The test plan provided appears to expand the scope of applicable data to the fuel form proposed to be tested. These proposed tests do not necessarily provide data in untested regimes but would provide assurance that the tested particles in the Xenergy fuel form meet fuel design acceptance criteria (as applicable to the final design). The NRC staff expects that an applicant would demonstrate, as part of a future submittal, the similarity of this tested fuel and recognizes the proposed testing program as a valid means to satisfy how the final fuel form is qualified, provided that the tests return acceptable results. This does not represent a limitation in the TR but does represent an item to be addressed as part of a future licensing application. The NRC staff notes that the tests performed are limited to the parameters of interest (e.g., temperature, burnup) documented in the TR. A final design submittal is expected to provide a full accounting of conditions that the fuel is subject to. These conditions may or may not affect the safety demonstration as described in the TR (for example, transient conditions for irradiated fuel). In the response to RAI question 4, Xenergy acknowledges additional parameters of interest that may impact the fuel qualification and will be identified in other future TRs or submittals and will be presented to the NRC staff for evaluation. The NRC staff findings are limited to the areas and conditions identified in the TR. The NRC staff considers this a limitation on the TR associated with Limitation 2. Diffusion of volatile metals has the potential to represent an important contributor to the source term for a fuel performance methodology as outlined in the TR (for the purposes of this discussion, volatile metals refers to those identified in Figure 33). The fuel qualification aspect of the TR focuses on intact particles, but these volatile metal radionuclides have the potential to diffuse through intact particles and contribute to the source term to be calculated by the fuel performance code. The NRC staff understands that this input is not necessarily within the scope of the TR as described, but the concept is adjacent and intertwined with the data output used by the tests described in the TR. Data from the testing may inform this input. As such, in addition to the manufacturing defects and contamination described above, applicants referencing the TR should explicitly consider diffusion radionuclides separate from the intact particle releases derived here. Combined with the discussion above in the Fuel Design and Performance Requirements section of this SE, this represents Limitation 1 on the TR. Portions of the TR Not Addressed by this SE The focus of this SE is for the methodology intended to demonstrate that the fuel used in the Xe100 is qualified to specified standards and release limits. As such, this SE does not make findings associated with a number of TR sections, which are viewed by the NRC staff as being provided for background and not related to the TR review objectives in Section 7. The sections of the TR not addressed by this SE include: Section 3.3, NRC Guidance/References
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Section 3.4, Additional Guidance: ANS 53.1 Section 3.5, US HTGR Precedents Section 5.2, TRISO-Coated Particle Fuel Performance in Spherical Fuel Elements (to the extent that results themselves are not being relied on; discussion of some points highlighted in this section are in the above evaluation) Appendix A, NRC Assessment of NGNP White Papers: Follow Up Items for Fuel Qualification Considerations related to fuel pebble mechanical effects and interactions with external influences were not discussed in the TR. Additionally, possible wear and tritium uptake were determined to be outside the scope of the TR, per the response to RAI question 4. The NRC staff anticipates that these would be the topic of future submittals and would be reviewed as relevant and appropriate at the time of the submittal. LIMITATIONS An applicant may reference the TR for use as applied to the applicants facility only if the applicant demonstrates compliance with the following limitations: 1. Applicants referencing the TR should provide a justification for releases from defect particles, contamination in the fuel matrix, and diffusion of volatile radionuclides in addition to the intact particle failure fraction derived from the testing to be used in a future licensing application. In effect, this methodology forms a portion of the fuel qualification for the final reactor fuel form, but not the entirety of the fuel qualification process for the performance of the fuel (e.g., cover the aspects cited above or the more comprehensive list of fuel qualification areas in NUREG-2246). 2. This approval is limited to the plan to test TRISO fuel and the operational envelope (temperature, burnup, and environmental conditions) outlined in the TR. Applicants referencing the TR should provide a set of fuel specifications and demonstrate how fuel to be used in the reactor will meet these specifications. Future licensing submittals will be needed to qualify the fuel for the totality of the operational regime, including but not limited to transient accident conditions and mechanical effects to the extent that these are not incorporated within the scope of the TR. 3. An applicant referencing the TR is subject to the limitations and conditions associated with the approved EPRI TRISO TR (TR References 2 and 7). CONCLUSION Based on the above discussion, the NRC staff concludes that the fuel qualification program, as described in the TR and subject to the limitations in this SE, provides an acceptable framework for qualifying fuel for the Xe100 reactor design. Specifically, the fuel design information developed as a result of the AGR program and augmented by the confirmatory testing proposed
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION in the TR can be used with other design information to establish a set of parameters, performance requirements, and a design basis envelope for qualified fuel for the Xe100 design. Further, the NRC staff concludes that the test plans identified in Section 6 provide for an acceptable pathway to demonstrate qualified performance of UCO TRISO for normal operation and steady-state accident heat-up conditions under the conditions described in Tables 19 and
- 20. Test results could be used in their entirety as part of a future licensing submittal to justify fuel performance under tested conditions.
As discussed in this SE, Xenergy indicated that the detailed design of some aspects of the Xe100 fuel design, and its relation to the final Xe100 design parameters, is not complete at the present time. The NRC staff will make a final determination of the Xe100 fuel design acceptability when the complete, detailed design, including applicable testing, is completed and reviewed by the NRC staff as part of future licensing activities referencing the TR. REFERENCES None. Principal Contributor: Boyce Travis Reed Anzalone Date: March 9, 2023
From: Maggie Staiger To: Kenny Nguyen Cc: ext Yvonne Mirowski; Adrian Muniz; Zackary Stone (He/Him/His)
Subject:
[External_Sender] TRISO-X Pebble Fuel Qualification Methodology LTR Date: Wednesday, September 6, 2023 8:47:20 AM Attachments: 2023-XE-NRC-018-A PUBLIC.pdf Hi Kenny, Please find the attached PUBLIC (non-proprietary /redacted) version of the Accepted Triso-X Pebble Fuel Qualification Methodology LTR. We are asking your assistance to replace the current PUBLIC document under ML23160A294 with the attached updated document. This is a portion of the documents that would have been submitted under submission ID# 41773 June 9th. Changes within this document: Removes the proprietary markings originally remaining on the PUBLIC Enclosure Verifies the pages are right side up. Thank you for all of your help these past months and have a great weekend. X-energy Maggie Staiger Licensing Project Manager / Engineer Mobile: mstaiger@x-energy.com}}