ML23130A208

Comment (9) of Mark A. Richter on Behalf of Nuclear Energy Institute on Material Compatibility for Non-Light Water Reactors

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 blob:https://www.fdms.gov/31306936-a00c-44cd-898b-940415b6d5e5 1/1 PUBLIC SUBMISSION As of: 5/10/23, 10:20 AM Received: May 08, 2023 Status: Pending_Post Tracking No. lhf-4x9e-62pj Comments Due: May 08, 2023 Submission Type: Web 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 SUNSI Review Complete Template=ADM-013 E-RIDS=ADM-03 ADD: Jordan Hoellman, Christopher Cauffman, Allen Hughes, Mary Neely Comment (9)

Publication Date: 3/7/2023 Citation: 88 FR 14186

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

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 it be acceptable to have materials from the same heat tested in a simulated environment?

Allow materials from the same heat to be tested in a simulated environment to satisfy the surveillance requirement.

General Several of the sections on degradation mechanisms end with the statement: The staff should confirm that applicants also consider appropriate mitigation strategies, performance monitoring, and surveillance programs to ensure that SSCs affected by corrosion continue to satisfy the design criteria.

It is suggested to be more specific on which design criteria the ISG refers to at the end, for example by saying to satisfy the principal design criteria.

Qualification and Performance Monitoring The following paragraph mentions testing, but it does not specifically allow for the use of data from previous facilities within the same parametric operating 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.

Guidance should be updated to allow for the use of data from previous facilities within the same parametric operating envelope.

Qualification and Performance Monitoring page 6 first paragraph The ISG says: In the meantime, staff should evaluate whether applicants have adequately addressed the following general degradation mechanisms for various reactor environments.

It is suggested that the extent to which such degradation mechanisms should be addressed should be commensurate with their safety significance. A possible rewording could be: In the meantime, 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 Comments on Draft DANU-ISG-2023-01 2

Section Comment/Basis Recommendation safety significance of the degradation mechanism.

Qualification and Performance Monitoring page 6 last paragraph When saying: Erosion products from SSCs have the potential for depositing elsewhere in the coolant flow path, affecting coolant flow patterns and local heat transfer properties, it is suggested to also say that these erosion products may undergo activation, thus contributing to the activity of the coolant itself.

A possible wording could be: In addition to potentially undergoing activation thus contributing to the coolants activation level, erosion products from SSCs have the potential.heat transfer properties.

Qualification and Performance Monitoring page 9 near end of first paragraph Correct typo: striping in place of stripping within the sentence: The staff should ensure that very high cycle fatigue due to thermal stripping has been adequately addressed by the applicant Qualification and Performance Monitoring end of Thermal and Fatigue Transients page 9 It is suggested to add whenever applicable within the sentence: The staff should verify that synergistic effects of thermal fatigue, vibratory fatigue, and creep-fatigue are addressed by the applicant so that it reads: The staff should verify that, whenever applicable, synergistic effects of thermal fatigue, vibratory fatigue, and creep-fatigue are addressed by the applicant Qualification and Performance Monitoring Wear/Fretting page 9 It is suggested to clarify the following paragraph:

The staff should consider the potential impacts of the specific coolant environment on wear and fretting, particularly in heat exchangers in steam generators. Depending on the reactor design, the interaction between the coolants in the primary, secondary, and steam generating loops may have adverse consequences for the reactor with regard to wear and fretting.

Specifically, in heat exchangers and in steam generators seem to be a duplication, and the latter can be deleted. Moreover, when speaking about interaction between coolants, it is not clear whether the subject interaction is between the coolants, or between the coolants and the heat exchanger structures. This part would benefit from a rewording.

Page 9, first paragraph under General Materials Issues The ISG says: The staff should evaluate whether applicants have adequately addressed the following design neutral materials issues as appropriate for the application and design. It is suggested to indicate that the extent with which such design-neutral material issues should be addressed should be commensurate with their safety significance.

A possible rewording could be: The staff should evaluate whether applicants have adequately addressed the following design-neutral materials issues as appropriate for the application and design, to an extent which should be commensurate with the safety significance of each issue.

Comments on Draft DANU-ISG-2023-01 3

Section Comment/Basis Recommendation Reactor-Specific Guidance, Part 1: Molten Salt Reactors Page 11 Graphite-salt compatibility considerations include fluorination of the graphite and formation of carbides (uranium carbide, chromium carbide, and others), as well as potential infiltration of molten salt into the graphite (ORNL, 2021a.)

Comment 1: ORNL, 2021a is missing in the draft ISG as a reference.

Comment 2: The fact that fluorination is 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 paragraph under Reactor-Specific Guidance, Part 1: Molten Salt Reactors It is recommended to correct the definition of MSRs operating with solid fuel, as the ISG indicates that in these MSRs the molten salt coolant has relatively small amounts of fissile material and fission products, which is not true as the fissile material is contained within the fuel, not the 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 Degradation Mechanisms -

Irradiation Guidance notes that: The staff should evaluate data on the effects of neutron irradiation on materials, including mechanisms such as irradiation-assisted creep, irradiation embrittlement, irradiation-assisted SCC, and decreased resistance to oxidation.

Given that III-5 does not provide specific acceptable means to account for irradiation effects on structural material properties, the guidance should be updated to provide additional detail on staff expectations for review or acceptable means to account for irradiation effects on structural material properties.

Comments on Draft DANU-ISG-2023-01 4

Section Comment/Basis Recommendation General Degradation Mechanisms:

Silicon Carbide (SiC)

1. SiC is captured in the General Materials Issues section with the main takeaway that all SiC types should be qualified separately.

2. For the reactor specific guidance sections, molten salt and liquid metal reactors (sodium and Lead coolant) both specifically call out SiC but strangely there is no SiC reference under HTGRs. Given this is where GA-EMS is most focused, perhaps we should look to add something in there.

a. Under the section Reactor Specific Guidance, Part 3: High Temperature Gas Reactor, the document states that NRC is not aware of any current plans to deploy GFR reactors in the Unites States, so this section does not address materials concerns for GFRs.

3. Not all of the degradation mechanisms are broadly applicable to all candidate reactor materials. For instance, stress relaxation cracking is an identified mechanism in heat affected zone of alloy welds, but is not an expected damage mechanism in silicon carbide material.

1. It should be acknowledged that degradation mechanisms are fundamentally different in SiC-SiC composites compared to metals (Jacobsen, GA-EMS, JNM, 452, p125-132, 2014). The staff should be aware that different mechanisms (e.g. - matrix cracking, fiber sliding) and different analytical techniques (e.g. - Weibull analysis) must be considered to account for the stochastic behavior of SiC materials.

2. This is an issue that needs to be resolved. GA-EMS is developing the Fast Modular Reactor with the intent to deploy in the US, and this design leverages non-metallic materials, specifically SiC/SiC composite due to its demonstrated high temperature performance and compatibility with Helium coolant (Choi, GA-EMS, ANS Transactions, 124, p454-456, 2021). As is written above in the molten salt and metal 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 Guidance, Part 1: Molten Salt Reactors Page 12 The staff should evaluate whether the application adequately addressed the potential formation of uranium and other metal carbides on graphite, and subsequent deleterious effects on reactor materials (EPRI, 2019a.)

Per the EPRI report, the only concern with metal carbide forming on graphite seems to be related to corrosion of metals, not related to degradation of 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.

Comments on Draft DANU-ISG-2023-01 5

Section Comment/Basis Recommendation Reactor-Specific Guidance, Part 2: Liquid Metal Reactors -

Caustic Stress-Corrosion Cracking Although it is noted that most steam generators, both tubes and shell, are made of ferritic steels, austenitic stainless steels have been used successfully in previous sodium fast reactor steam generators (e.g., EBR-II and the Prototype Fast Reactor (PFR) operated on the Dounreay site).

Guidance should be updated to reflect this operational experience.

Reactor-Specific Guidance, Part 2: Liquid Metal Reactors -

Impurity Effects on Corrosion Successfully operated sodium fast reactors (e.g.,

the Experimental Breeder Reactor-II) and standards developed for SFR systems have established maximum acceptable oxygen levels in sodium of 2 ppm.

EBR-II Operating Experience, Section 5.2, Source 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.

While it is noted that higher oxygen concentration has been seen to increase the corrosion rates of steels in a sodium environment, the guidance should be updated to reflect that oxygen levels of 2 ppm have been shown to be acceptable.

Page 13, first paragraph under Reactor-Specific Guidance, Part 2: Liquid Metal Reactors When introducing LBE it is suggested to indicate the composition of this eutectic, for example by adding it at the end of the sentence: Liquid metal reactors are characterized by their operation at or near ambient pressure using a fast neutron spectrum in which the fuel, with metallic cladding, is cooled by 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 paragraph under Reactor-Specific Guidance, Part 2: Liquid Metal Reactors When saying: To date, operational experience with LFRs is limited to propulsion nuclear reactors in Alfa-class submarines operated by the Soviet Union from 1967-1983, it is suggested to specify that these reactors were LBE-cooled.

A possible rewording could be: To date, operational experience with LFRs is limited to LBE-cooled propulsion nuclear reactors in Alfa-class submarines operated by the Soviet Union from 1967-1983.

Page 15, first sentence under Lead coolant section The ISG says: A lead-cooled reactor may use lead (Tmelt, 327.5 degrees C) or LBE alloy (Tmelt, 123.5 degrees C) as the coolant. As LBE has a significant (~55%) content in bismuth, it is suggested not to refer to the corresponding reactor as lead-cooled.

A proposed rewording could be:

Reactors operating with lead-based coolants may use lead (Tmelt, 327.5 degrees C) or LBE alloy (Tmelt, 123.5 degrees C) as the coolant.

Comments on Draft DANU-ISG-2023-01 6

Section Comment/Basis Recommendation Page 15, first paragraph under Lead coolant section The ISG says: As a result, use of typical ferritic and austenitic steels requires special treatments, such as alloying additions or coatings (EPRI, 2019b).

It is recommended to correct this sentence, as the need for special treatments is not absolute but depends 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 first paragraph under Lead coolant section It is suggested to more strongly emphasize the (correct) statement: Specific data of the environmental impacts of molten lead and LBE on 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 paragraph under Lead coolant section The ISG says: The staff should evaluate whether applicants have adequately addressed the following materials issues, including plans to monitor, evaluate, and mitigate degradation. It is suggested to add that the extent to which this is addressed should be commensurate with the safety significance of the degradation mechanism.

A proposed rewording could be: The staff should evaluate whether applicants have adequately addressed the following materials issues, including plans to monitor, evaluate, and mitigate degradation, in a way commensurate with the safety significance associated with each degradation mechanism.

Page 15, last paragraph It is suggested to add an indication of the 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:

Comments on Draft DANU-ISG-2023-01 7

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 on Lead erosion It is recommended to reword the statement: Lead is highly eroding, and for this reason, the flow velocity should be limited (IRSN, 2012; Ballinger and Lim, 2003) as neither of the references provided gives evidence that lead is highly eroding.. While it is true that the lead flow velocity must be limited to prevent erosion effects, as written the text is misleading.

It is suggested to reword as: At high lead (or LBE) flow velocities the effect of erosion, in addition to corrosion, should be considered. Even though the flow velocity limit is not absolute but temperature-and material-dependent, common practice is to maintain the 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 Comments on Draft DANU-ISG-2023-01 8

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 to last paragraph within section on Liquid Metal Embrittlement The ISG says: The severity of dissolution corrosion attack in Type 316L stainless steel was found to increase with increasing percentages of cold work (Klok et al., 2017). This is evidence collected in LBE coolant, which is known to behave differently than pure lead.

As such, it is suggested to reword the sentence by saying: The severity of dissolution corrosion attack in Type 316L stainless steel was found, in LBE coolant, to increase with increasing percentages of cold work (Klok et al.,

2017).

Page 16, section on Nonmetallic materials The ISG says: SiC/SiC has shown resistance to liquid metal corrosion up to 800 degrees C in a few short-term tests using a non-flowing lead-lithium eutectic (EPRI, 2019b).

It is recommended not to use this example/reference as it is referred to a coolant, i.e., lead-lithium eutectic, that 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-Comments on Draft DANU-ISG-2023-01 9

Section Comment/Basis Recommendation 1065. http://refhub.elsevier.com/B978-0-12-803581-8.00749-9/sbref146 Page 16, bottom of the page under Oxygen control The ISG says: Unlike in other reactor types, accelerated corrosion can occur if the dissolved oxygen concentration is either too high or too low at a specific temperature (EPRI, 2019b; Klok et al.,

2018). This statement is not correct since, for example, if the oxygen content is very high and the temperature is below 450-480°C there will not be any significant corrosion (there will be, however, precipitation of lead oxide which should not be confused with corrosion).

In light of this, it is suggested to replace that statement with the following:

Oxygen control is an important technique to ensure satisfactory performance of structural materials in lead-and LBE-cooled reactors. This technique, widely used in lead-based test facilities worldwide (Ref. A),

consists of maintaining the oxygen 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 Comments on Draft DANU-ISG-2023-01 10 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 Manufacturing Technologies The staff should evaluate whether an application containing AMT components considers (1) the differences between the AMT and traditional manufacturing methods; (2) the safety significance of the identified differences; (3) the aspects of each AMT that are not currently addressed by codes and standards or regulations; and (4) the impacts of the proposed reactor type, operating conditions, and material on the AMT qualification and performance.

The staff should confirm that applicants also consider appropriate mitigation strategies, performance monitoring, and surveillance programs to ensure that SSCs fabricated by AMTs continue to satisfy the design criteria.

The guidance should be updated to remove the study of differences between AMT and traditional technologies. Technology is not less safe because it is newer. Likewise, safety is not affected by technology being different. Advanced manufacturing technologies should be evaluated on their own merits and the products of those technological methods. This is most likely what was meant, but clarification is needed.