ML23130A208
| ML23130A208 | |
| Person / Time | |
|---|---|
| Site: | Nuclear Energy Institute |
| Issue date: | 05/08/2023 |
| From: | Richter M Nuclear Energy Institute |
| To: | NRC/SECY, Office of Administration |
| References | |
| NRC-2022-0215, 88FR14186 00009 | |
| Download: ML23130A208 (1) | |
Text
5/10/23, 10:20 AM blob:https://www.fdms.gov/31306936-a00c-44cd-898b-940415b6d5e5 SUNSI Review Complete Template=ADM-013 As of: 5/10/23, 10:20 AM E-RIDS=ADM-03 Received: May 08, 2023 PUBLIC SUBMISSION ADD: Jordan Hoellman, Christopher Cauffman, Allen Status: Pending_Post Tracking No. lhf-4x9e-62pj Hughes, Mary Neely Comment (9) Comments Due: May 08, 2023 Publication Date: 3/7/2023 Submission Type: Web Citation: 88 FR 14186 Docket: NRC-2022-0215 Material Compatibility Interim Staff Guidance Comment On: NRC-2022-0215-0001 Material Compatibility for Non-Light Water Reactors Document: NRC-2022-0215-DRAFT-0009 Comment on FR Doc # 2023-04577 Submitter Information Email: atb@nei.org Organization: Nuclear Energy Institute General Comment See attached file(s)
Attachments 05-08-23_NRC_NEI Comment Letter_DANU-ISG-2023-01_Attach blob:https://www.fdms.gov/31306936-a00c-44cd-898b-940415b6d5e5 1/1
MARK A. RICHTER, PH.D.
Technical Advisor, Decommissioning & Used Fuel 1201 F Street, NW, Suite 1100 Washington, DC 20004 P: 202.739.8106 mar@nei.org nei.org May 8, 2023 Office of the Secretary U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Attn: Program Management, Announcements and Editing Staff Submitted on Regulations.gov
Subject:
NEI Comments Regarding NRC Draft Interim Staff Guidance (ISG) Material Compatibility for non-Light Water Reactors, DANU-ISG-2023-01 (Docket ID NRC-2022-0215)
Project Number: 689
Dear Program Management,
Announcements and Editing Staff:
On behalf of the Nuclear Energy Institutes (NEI) 1 members, we are grateful for the opportunity to review and offer comments regarding NRC Draft Interim Staff Guidance (ISG), Material Compatibility for non-Light Water Reactors, DANU-ISG-2023-01 (Docket ID NRC-2022-0215.) The purpose of this draft ISG is to assist the NRC staff in reviewing certain applications for construction and operation of non-light water reactor designs, including power and non-power reactors.
NEI offers specific comments on DANU-ISG-2023-01 related to degradation mechanisms of materials proposed for use in non-light water reactors, such as corrosion and radiation effects, in the attached comment table.
1 The Nuclear Energy Institute (NEI) is responsible for establishing unified policy on behalf of its members relating to matters affecting the nuclear energy industry, including the regulatory aspects of generic operational and technical issues. NEIs members include entities licensed to operate commercial nuclear power plants in the United States, nuclear plant designers, major architect and engineering firms, fuel cycle facilities, nuclear materials licensees, and other organizations involved in the nuclear energy industry.
Thank you for your time and consideration of the attached comments. If you have any questions, please contact me.
Sincerely, Mark A. Richter Attachment cc: Ron Faibish, General Atomics Paolo Ferroni, Westinghouse Electric Company LLC Ross Moore, Oklo John Price, Kairos Power
Attachment 1 Comments on Draft DANU-ISG-2023-01, Material Compatibility for non-Light Water Reactors Section Comment/Basis Recommendation General For performance monitoring and surveillance, would Allow materials from the same heat to it be acceptable to have materials from the same be tested in a simulated environment to heat tested in a simulated environment? satisfy the surveillance requirement.
General Several of the sections on degradation mechanisms It is suggested to be more specific on end with the statement: The staff should confirm which design criteria the ISG refers to that applicants also consider appropriate mitigation at the end, for example by saying to strategies, performance monitoring, and satisfy the principal design criteria.
surveillance programs to ensure that SSCs affected by corrosion continue to satisfy the design criteria.
Qualification The following paragraph mentions testing, but it Guidance should be updated to allow for and does not specifically allow for the use of data from the use of data from previous facilities Performance previous facilities within the same parametric within the same parametric operating Monitoring operating envelope. envelope.
Materials qualification and monitoring programs should include testing conducted in an environment simulating the anticipated operating environment for the reactor, including chemical environment, temperatures, and irradiation. Testing should account for uncertainties in the environment, material composition, fabrication methods, and operating conditions. The scope of this testing should include safety-related component materials, safety-significant component materials, and as needed, non-safety related component materials whose failure could impact critical design functions.
Testing should be conducted to determine if materials properties and allowable stresses meet applicable codes and standards or other design requirements. If necessary, appropriate reduction factors should be applied to the materials properties and allowable stresses from the applicable design codes and/or design specifications.
Qualification The ISG says: In the meantime, staff should It is suggested that the extent to which and evaluate whether applicants have adequately such degradation mechanisms should Performance addressed the following general degradation be addressed should be commensurate Monitoring page mechanisms for various reactor environments. with their safety significance. A possible 6 first rewording could be: In the meantime, paragraph staff should evaluate whether applicants have adequately addressed the following general degradation mechanisms for various reactor environments, to an extent which should be commensurate with the 1
Comments on Draft DANU-ISG-2023-01 Section Comment/Basis Recommendation safety significance of the degradation mechanism.
Qualification When saying: Erosion products from SSCs have the A possible wording could be: In and potential for depositing elsewhere in the coolant addition to potentially undergoing Performance flow path, affecting coolant flow patterns and local activation thus contributing to the Monitoring page heat transfer properties, it is suggested to also say coolants activation level, erosion 6 last paragraph that these erosion products may undergo activation, products from SSCs have the thus contributing to the activity of the coolant itself. potential.heat transfer properties.
Qualification Correct typo: striping in place of stripping within and the sentence: The staff should ensure that very Performance high cycle fatigue due to thermal stripping has been Monitoring page adequately addressed by the applicant 9 near end of first paragraph Qualification It is suggested to add whenever applicable within and the sentence: The staff should verify that Performance synergistic effects of thermal fatigue, vibratory Monitoring end fatigue, and creep-fatigue are addressed by the of Thermal and applicant so that it reads: The staff should verify Fatigue that, whenever applicable, synergistic effects of Transients thermal fatigue, vibratory fatigue, and creep-fatigue page 9 are addressed by the applicant Qualification It is suggested to clarify the following paragraph: Specifically, in heat exchangers and and The staff should consider the potential impacts of in steam generators seem to be a Performance the specific coolant environment on wear and duplication, and the latter can be Monitoring fretting, particularly in heat exchangers in steam deleted. Moreover, when speaking Wear/Fretting generators. Depending on the reactor design, the about interaction between coolants, it page 9 interaction between the coolants in the primary, is not clear whether the subject secondary, and steam generating loops may have interaction is between the coolants, or adverse consequences for the reactor with regard between the coolants and the heat to wear and fretting. exchanger structures. This part would benefit from a rewording.
Page 9, first The ISG says: The staff should evaluate whether A possible rewording could be: The paragraph applicants have adequately addressed the following staff should evaluate whether applicants under General design neutral materials issues as appropriate for have adequately addressed the Materials the application and design. It is suggested to following design-neutral materials issues Issues indicate that the extent with which such design- as appropriate for the application and neutral material issues should be addressed should design, to an extent which should be be commensurate with their safety significance. commensurate with the safety significance of each issue.
2 Comments on Draft DANU-ISG-2023-01 Section Comment/Basis Recommendation Reactor-Specific Graphite-salt compatibility considerations include Comment 1: ORNL, 2021a is missing in Guidance, Part fluorination of the graphite and formation of the draft ISG as a reference.
1: Molten Salt carbides (uranium carbide, chromium carbide, and Reactors Page others), as well as potential infiltration of molten Comment 2: The fact that fluorination is 11 salt into the graphite (ORNL, 2021a.) the first stated compatibility issue might be concerning. There is no relevant data that fluorination is a real thing for all engineering purposes (although it is mentioned in the literature). At most, this should be demonstrated with a test program and not require monitoring.
Comment 3: Formation of carbides is not a concern for the graphite itself (see comment 1 below for Page 12).
Comment 4: Infiltration should be demonstrated with a test program and not require monitoring. Dispensing with monitoring might require more data from the test program to show its not happening and/or effects are very mild.
Its a bigger concern for fuel salt MSRs because the salt pressure is higher (salt is denser), hot spots, and Xe accumulation.
Page 11, first It is recommended to correct the definition of MSRs paragraph operating with solid fuel, as the ISG indicates that under Reactor- in these MSRs the molten salt coolant has Specific relatively small amounts of fissile material and Guidance, Part fission products, which is not true as the fissile 1: Molten Salt material is contained within the fuel, not the Reactors coolant. In addition, when referring to TRISO later in the same paragraph, it is suggested to indicate that, although dominant, this is just an example of solid fuel used in solid-fuel MSRs.
General Guidance notes that: The staff should evaluate Given that III-5 does not provide Degradation data on the effects of neutron irradiation on specific acceptable means to account for Mechanisms - materials, including mechanisms such as irradiation- irradiation effects on structural material Irradiation assisted creep, irradiation embrittlement, properties, the guidance should be irradiation-assisted SCC, and decreased resistance updated to provide additional detail on to oxidation. staff expectations for review or acceptable means to account for irradiation effects on structural material properties.
3 Comments on Draft DANU-ISG-2023-01 Section Comment/Basis Recommendation General 1. SiC is captured in the General Materials Issues 1. It should be acknowledged that Degradation section with the main takeaway that all SiC degradation mechanisms are Mechanisms: types should be qualified separately. fundamentally different in SiC-SiC Silicon Carbide composites compared to metals
- 2. For the reactor specific guidance sections, (SiC) (Jacobsen, GA-EMS, JNM, 452, molten salt and liquid metal reactors (sodium p125-132, 2014). The staff should and Lead coolant) both specifically call out SiC be aware that different mechanisms but strangely there is no SiC reference under (e.g. - matrix cracking, fiber sliding)
HTGRs. Given this is where GA-EMS is most and different analytical techniques focused, perhaps we should look to add (e.g. - Weibull analysis) must be something in there.
considered to account for the
- a. Under the section Reactor Specific stochastic behavior of SiC materials.
Guidance, Part 3: High Temperature Gas
- 2. This is an issue that needs to be Reactor, the document states that NRC is resolved. GA-EMS is developing the not aware of any current plans to deploy Fast Modular Reactor with the GFR reactors in the Unites States, so this intent to deploy in the US, and this section does not address materials concerns design leverages non-metallic for GFRs.
materials, specifically SiC/SiC
- 3. Not all of the degradation mechanisms are composite due to its demonstrated broadly applicable to all candidate reactor high temperature performance and materials. For instance, stress relaxation compatibility with Helium coolant cracking is an identified mechanism in heat (Choi, GA-EMS, ANS Transactions, affected zone of alloy welds, but is not an 124, p454-456, 2021). As is written expected damage mechanism in silicon carbide above in the molten salt and metal material. coolant sections, the staff should be aware of the potential sources and impacts of impurities in the Helium and the effects these have on SiC/SiC performance and degradation mechanisms.
- 3. There should be an avenue to specify if certain mechanisms arent applicable to a material system or plant design, in addition to the stated feasibility of adding additional mechanisms.
Reactor-Specific The staff should evaluate whether the application Per the EPRI report, the only concern Guidance, Part adequately addressed the potential formation of with metal carbide forming on graphite 1: Molten Salt uranium and other metal carbides on graphite, and seems to be related to corrosion of Reactors Page subsequent deleterious effects on reactor materials metals, not related to degradation of 12 (EPRI, 2019a.) the graphite itself. Also, the concern with uranium carbide for fuel-salt MSRs is related to nuclear performance (neutronics), not graphite degradation.
This evaluation should be removed as a compatibility issue.
4 Comments on Draft DANU-ISG-2023-01 Section Comment/Basis Recommendation Reactor-Specific Although it is noted that most steam generators, Guidance should be updated to reflect Guidance, Part both tubes and shell, are made of ferritic steels, this operational experience.
2: Liquid Metal austenitic stainless steels have been used Reactors - successfully in previous sodium fast reactor steam Caustic Stress- generators (e.g., EBR-II and the Prototype Fast Corrosion Reactor (PFR) operated on the Dounreay site).
Cracking Reactor-Specific Successfully operated sodium fast reactors (e.g., While it is noted that higher oxygen Guidance, Part the Experimental Breeder Reactor-II) and standards concentration has been seen to increase 2: Liquid Metal developed for SFR systems have established the corrosion rates of steels in a sodium Reactors - maximum acceptable oxygen levels in sodium of 2 environment, the guidance should be Impurity Effects ppm. updated to reflect that oxygen levels of on Corrosion 2 ppm have been shown to be EBR-II Operating Experience, Section 5.2, Source acceptable.
Rate of Impurities, (1978) notes that EBR-II operating limits for primary sodium are 2.0 ppm oxygen and 200 ppb hydrogen. Normal concentrations are ~0.8 ppm oxygen and ~90 ppb hydrogen. RDT A 1-5T, Purity Requirements for Operating Sodium Reactor Systems, (1973) specifies an oxygen concentration limit of up to 2.0 ppm for hot leg temperatures >800 F.
Page 13, first When introducing LBE it is suggested to indicate the paragraph composition of this eutectic, for example by adding under Reactor- it at the end of the sentence: Liquid metal reactors Specific are characterized by their operation at or near Guidance, Part ambient pressure using a fast neutron spectrum in 2: Liquid Metal which the fuel, with metallic cladding, is cooled by Reactors liquid sodium, lead, or the lead-bismuth eutectic (LBE, 44.5 wt% Pb and 55.5 wt% Bi). This is to clarify that the composition of this eutectic is very far from pure Pb.
Page 13, first When saying: To date, operational experience with A possible rewording could be: To paragraph LFRs is limited to propulsion nuclear reactors in date, operational experience with LFRs under Reactor- Alfa-class submarines operated by the Soviet Union is limited to LBE-cooled propulsion Specific from 1967-1983, it is suggested to specify that nuclear reactors in Alfa-class Guidance, Part these reactors were LBE-cooled. submarines operated by the Soviet 2: Liquid Metal Union from 1967-1983.
Reactors Page 15, first The ISG says: A lead-cooled reactor may use A proposed rewording could be:
sentence under lead (Tmelt, 327.5 degrees C) or LBE alloy (Tmelt, Reactors operating with lead-based Lead coolant 123.5 degrees C) as the coolant. As LBE has a coolants may use lead (Tmelt, 327.5 section significant (~55%) content in bismuth, it is degrees C) or LBE alloy (Tmelt, 123.5 suggested not to refer to the corresponding reactor degrees C) as the coolant.
as lead-cooled.
5 Comments on Draft DANU-ISG-2023-01 Section Comment/Basis Recommendation Page 15, first The ISG says: As a result, use of typical ferritic and It is recommended to correct this paragraph austenitic steels requires special treatments, such sentence, as the need for special under Lead as alloying additions or coatings (EPRI, 2019b). treatments is not absolute but depends coolant section on the temperature. Specifically, typical steels do not require special treatments when the temperature is below approx 480C, which is the operating temperature for internals operating at cold leg temperature. A possible rewording could be: As a result, when the temperature is above approximately 480°C, use of typical ferritic and austenitic steels requires special treatments, such as alloying additions or coatings (EPRI, 2019b)
Page 15, end of It is suggested to more strongly emphasize the first paragraph (correct) statement: Specific data of the under Lead environmental impacts of molten lead and LBE on coolant section materials are not interchangeable, as the two are often confused. A proposed rewording is: It should be stressed that specific data on the environmental impact of molten lead and LBE on materials are not interchangeable since, for the same temperature, LBE is typically more corrosive than pure lead (Ref.
X). where Ref. X is: NEA-OECD, Handbook on Lead-bismuth Eutectic Alloy and Lead Properties, Materials Compatibility, Thermal-hydraulics and Technologies. 2015 edition Page 15, second The ISG says: The staff should evaluate whether A proposed rewording could be: The paragraph applicants have adequately addressed the following staff should evaluate whether applicants under Lead materials issues, including plans to monitor, have adequately addressed the coolant section evaluate, and mitigate degradation. It is suggested following materials issues, including to add that the extent to which this is addressed plans to monitor, evaluate, and mitigate should be commensurate with the safety degradation, in a way commensurate significance of the degradation mechanism. with the safety significance associated with each degradation mechanism.
Page 15, last It is suggested to add an indication of the paragraph temperature range and additional references at the end of the following sentence:
Non-code-qualified materials such as alumina-forming or aluminum-coated stainless steels and silicon-enriched stainless steels may provide enhanced corrosion resistance in LBE and lead coolants at high temperatures (EPRI, 2019b; OECD, 2007; Ballinger and Lim, 2003), so that it reads:
6 Comments on Draft DANU-ISG-2023-01 Section Comment/Basis Recommendation Non-code-qualified materials () in LBE and lead coolants at high temperatures up to at least 700-750°C (EPRI, 2019b; OECD, 2007; Ballinger and Lim, 2003; Ref. A, Ref. B, Ref. C, Ref. D, Ref. E) where the references are: Ref. A: F. García Ferré, et al., Corrosion and radiation resistant nanoceramic coatings for lead fast reactors, Corrosion Science, 124 (2017) 80-92.
Ref. B: DOMSTEDT, P., LUNDBERG, M., SZAKALOS, P., Corrosion Studies of Low-Alloyed FeCrAl Steels in Liquid Lead at 750 °C. Oxidation of Metals (2019) 91:511-524. https://doi.org/10.1007/s11085-019-09896-z Ref. C: DOMSTEDT, P., et. al., (2020), Corrosion studies of a low alloyed Fe-10Cr-4Al steel exposed in liquid Pb at very high temperatures. Journal of Nuclear Materials. 531. 152022.
10.1016/j.jnucmat.2020.152022.
Ref. D: CHEN, L., et al., Investigation of microstructure and liquid lead corrosion behavior of a Fe-18Ni-16Cr-4Al base alumina-forming austenitic stainless steel. Mater. Res. Express 7 (2020) 026533. https://doi.org/10.1088/2053-1591/ab71d1 Ref. E: PINT, B.A., SU, Y.F., BRADY, M.P., et. al.,
Compatibility of Alumina-Forming Austenitic Steels in Static and Flowing Pb. JOM 73, 4016-4022 (2021). https://doi.org/10.1007/s11837-021-04961-y Page 16, section It is recommended to reword the statement: Lead It is suggested to reword as: At high on Lead is highly eroding, and for this reason, the flow lead (or LBE) flow velocities the effect erosion velocity should be limited (IRSN, 2012; Ballinger of erosion, in addition to corrosion, and Lim, 2003) as neither of the references should be considered. Even though the provided gives evidence that lead is highly flow velocity limit is not absolute but eroding.. While it is true that the lead flow velocity temperature- and material-dependent, must be limited to prevent erosion effects, as common practice is to maintain the written the text is misleading. velocity of lead-based coolants in high-temperature regions of the reactor coolant system, such as the core, below approximately 2 m/s both for LBE (Ref.
A) and for pure lead coolant (Ref. B).
Higher velocities may be acceptable especially when the operating temperature is low, such as for pump impellers located in the cold leg of the 7
Comments on Draft DANU-ISG-2023-01 Section Comment/Basis Recommendation reactor coolant system where the temperature is generally at or below 400°C. where the references are:
Ref. A: T. R. Allen and D. C. Crawford, Lead-Cooled Fast Reactor Systems and the Fuels and Materials Challenges.
Science and Technology of Nuclear Installations, Volume 2007, Article ID 97486, doi:10.1155/2007/97486 Ref. B: Vogt, J.-B.; Proriol Serre, I. A Review of the Surface Modifications for Corrosion Mitigation of Steels in Lead and LBE. Coatings 2021, 11, 53. https://
doi.org/10.3390/coatings11010053 Page 16, second The ISG says: The severity of dissolution corrosion As such, it is suggested to reword the to last attack in Type 316L stainless steel was found to sentence by saying: The severity of paragraph increase with increasing percentages of cold work dissolution corrosion attack in Type within section (Klok et al., 2017). This is evidence collected in 316L stainless steel was found, in LBE on Liquid Metal Embrittlement LBE coolant, which is known to behave differently coolant, to increase with increasing than pure lead. percentages of cold work (Klok et al.,
2017).
Page 16, section The ISG says: SiC/SiC has shown resistance to It is recommended not to use this on Nonmetallic liquid metal corrosion up to 800 degrees C in a few example/reference as it is referred to a materials short-term tests using a non-flowing lead-lithium coolant, i.e., lead-lithium eutectic, that eutectic (EPRI, 2019b). is not relevant for fission applications and is completely different from the corrosion standpoint with respect to Pb and LBE. The following statement is suggested: SiC/SiC has shown resistance to liquid metal corrosion up to 550 degrees C in 2000 hr corrosion tests in flowing liquid LBE (Ref. A).
where the reference is:
Ref. A: TAKAHASHI, M., KONDO, M.,
Corrosion resistance of ceramics SiC and Si3N4 in flowing lead-bismuth eutectic.
Progress in Nuclear Energy, Volume 53, Issue 7, September 2011, Pages 1061-8 Comments on Draft DANU-ISG-2023-01 Section Comment/Basis Recommendation 1065. http://refhub.elsevier.com/B978-0-12-803581-8.00749-9/sbref146 Page 16, bottom The ISG says: Unlike in other reactor types, In light of this, it is suggested to replace of the page accelerated corrosion can occur if the dissolved that statement with the following:
under Oxygen oxygen concentration is either too high or too low Oxygen control is an important control at a specific temperature (EPRI, 2019b; Klok et al., technique to ensure satisfactory 2018). This statement is not correct since, for performance of structural materials in example, if the oxygen content is very high and the lead- and LBE-cooled reactors. This temperature is below 450-480°C there will not be technique, widely used in lead-based any significant corrosion (there will be, however, test facilities worldwide (Ref. A),
precipitation of lead oxide which should not be consists of maintaining the oxygen confused with corrosion). concentration in the coolant within technical specifications, which are generally represented by a minimum required concentration (needed to form a stable passivating oxide layer protecting the underlying bulk material) and a maximum allowed concentration (above which precipitation of lead oxide would occur thus causing plugging concerns for narrow flow passages).
The width of this permissible oxygen concentration window is a function of the reactors operating temperatures (both cold and hot leg temperatures),
with the lower end of this window (i.e.,
minimum required concentration) dependent on the corrosion protection technique leveraged by the class of materials used in the reactor coolant system. For example, adoption of conventional steels such as 316H would rely on the formation of a passivating iron oxide layer which requires a minimum oxygen concentration of approximately 10-8 wt% at 500°C.
Relaxation of this limit, thus permitting lower oxygen concentrations thus easing requirements on the oxygen control system, can be achieved by 9
Comments on Draft DANU-ISG-2023-01 Section Comment/Basis Recommendation adopting protective coatings/layers either artificially deposited on (or superficially diffused into) the components before they enter service, e.g., aluminization through pack cementation or Al2O3 deposition through Pulsed Laser Deposition, or self-forming/regenerating on the surface of certain materials, e.g.,
Alumina-Forming Austenitic steels (Ref.
B) where the references are:
Ref. A: M. Tarantino, et al., Overview on Lead-Cooled Fast Reactor Design and Related Technologies Development in ENEA. Energies 2021, 14, 5157.
https://doi.org/10.3390/en141651 Ref. B: S. BASSINI et al., Material Performance in Lead, in Comprehensive Nuclear Materials, Vol.
4, 2nd ed., L.- B. ALLOY, R. J. M.
KONINGS, and E. STOLLER ROGER, Eds. pp. 218-241, Elsevier, Oxford (2020).
Advanced The staff should evaluate whether an application The guidance should be updated to Manufacturing containing AMT components considers (1) the remove the study of differences Technologies differences between the AMT and traditional between AMT and traditional manufacturing methods; (2) the safety significance technologies. Technology is not less of the identified differences; (3) the aspects of each safe because it is newer. Likewise, AMT that are not currently addressed by codes and safety is not affected by technology standards or regulations; and (4) the impacts of the being different. Advanced proposed reactor type, operating conditions, and manufacturing technologies should be material on the AMT qualification and performance. evaluated on their own merits and the The staff should confirm that applicants also products of those technological consider appropriate mitigation strategies, methods. This is most likely what was performance monitoring, and surveillance programs meant, but clarification is needed.
to ensure that SSCs fabricated by AMTs continue to satisfy the design criteria.
10