ML23062A728

Redacted Final Safety Evaluation of Metallic Material Qualification for the Kairos Power Fluoride Salt-Cooled High-Temperature Reactor (KP-TR-013)

ML23062A728
Person / Time
Site:	99902069, Hermes  File:Kairos Power icon.png
Issue date:	03/13/2023
From:	Jessup W NRC/NRR/DANU/UAL1
To:	Hastings P Kairos Power
Rivera R
References
EPID L-2020-TOP-0050, CAC 000431
Download: ML23062A728 (1)

Text

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March 13, 2023

FINAL SAFETY EVALUATION OF METALLIC MATERIAL QUALIFICATION FOR THE KAIROS POWER FLUORIDE SALT-COOLED HIGH-TEMPERATURE REACTOR (KP-TR-013) KAIROS POWER, LLC EPID NO. 000431 / 99902069 / L-2020-TOP-0050

1.0 SPONSOR INFORMATION

Sponsor: Kairos Power, LLC (Kairos)

Address: 707 West Tower Ave.

Alameda, CA 94501

Project No.: 99902069 (Construction Permit Application Docket No. 05007513)

2.0 SUBMITTAL, CORRESPONDENCE, AND CONTRIBUTORS

2.1. Submittal Information

Revision 0 June 30, 2020 ML20182A799 KP-TR-014, Revision 0 Revision 1 June 30, 2021 ML21181A385 KP-TR-014, Revision 1 Revision 2 April 2, 2022 ML22116A246 KP-TR-014, Revision 2 Revision 3 August 19, 2022 ML22231B221 KP-TR-014, Revision 3 Revision 4 September 20, 2022 ML22263A456 KP-TR-014, Revision 4

• Agencywide Documents Access and Managem ent System (ADAMS) Accession No.

2.2. NRC Correspondence and Communications

Communication Type Date ADAMS Accession No.

Acceptance Review(s): September 3, 2020 ML20224A172 Closed Meeting Notices: December 6, 2021 ML21336A400 February 3, 2022 ML22032A336 February 14, 2022 ML22032A336 July 18, 2022 ML22196A385 August 10, 2022 ML22214A131 September 12, 2022 ML22244A250

• ADAMS Accession No.

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2.3. Principal Contributor(s)

John Honcharik, NRR/DNRL/NPHP Alexander Chereskin, NRR/DANU/UTB2 Richard Rivera, NRR/DANU/UAL1

3.0 BRIEF DESCRIPTION OF REQUEST AND BACKGROUND

Kairos Power, LLC (Kairos, the sponsor) is requesting Nuclear Regulatory Commission (NRC) staff review and approval of topical report (TR) KP-TR-013, Metallic Materials Qualification for the Kairos Power Fluoride Salt-Cooled, High Temperature Reactor, Revision 2, dated April 2022. The TR could apply to reactors using the Kairos Power Fluoride Salt-Cooled High Temperature Reactor (KP-FHR) designs 1 and could be used to support future licensing actions under Title 10 of Code of Federal Regulations (10 CFR) Parts 50 or 52. The TR includes the qualification plan for metallic structural materials used in Flibe-wetted areas for safety-related high temperature components of the KP-FHR power and non-power (test) reactors. Kairos also requested NRC approval of the planned material testing and analyses to address the materials reliability and compatibility in the environment of the KP-FHR designs. The results of these planned tests and analyses will be provided in a future license application that references this TR, along with a detailed description of the design, inspection, and surveillance programs for the KP-FHR designs.

The documents located at the ADAMS Accession number(s) identified in Section 2 of this SE have additional details on the submittal.

4.0 EVALUATION CRITERIA

4.1 Regulatory Requirements

The information Kairos will gather through their metallic material qualification program will satisfy, in part, 10 CFR 50.10, 10 CFR 50.34, 10 CFR 52.47, 10 CFR 52.79, 10 CFR 52.137, 10 CFR 52.157, which describe the requirements for the content of applications of limited work authorizations, construction permits, operating licenses, design certifications, combined licenses, standard design approvals, and manufacturing licenses, respectively.

4.2 Principal Design Criteria for the KP-FHR, Approved by the NRC Staff

The topical report KP-TR-003-P-A, Principal Design Criteria (PDC) for the Kairos Power Fluoride Salt-Cooled High Temperature Reactor, Revision 1, dated May 2020, provides PDCs for the KP-FHR design that were reviewed and approved by the NRC staff. The PDCs below are applicable to qualification of metallic components for the KP-FHR designs.

KP PDC 14, Reactor coolant boundary, which requires safety significant elements of the reactor coolant boundary to have an extremely low probability of abnormal leakage, rapidly propagating failure, or gross rupture. The continued performance of high temperature structural materials and the associated corrosion within the coolant relate to PDC 14.

1 When the term KP-FHR designs is referenced in this safety evaluation (SE), it applies to both the power reactor and non-power test reactor, unless otherwise specified.

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KP PDC 31, Fracture prevention of reactor coolant boundary, which requires, in part, the reactor coolant boundary to behave in a nonbrittle manner and to minimize the probability of rapidly propagating failure of the reactor coolant boundary, accounting for effects of coolant composition on material properties. The design refl ects consideration of service temperatures, service degradation of material properties, creep, fatigue, and other conditions of the boundary material under operating, maintenance, testing, and postulated accident conditions, and the uncertainties in determining: (1) material properties, (2) the effects of irradiation and coolant composition, including contaminants and reaction products, on material properties, (3) residual, steady state, and transient stresses, and (4) size of flaws.

4.3 Codes, Standards, and Guidance Documents

Applicable Codes and Standards:

The NRC staff also considered the following codes and standards and guidance documents during the course of its review:

American Society of Mechanical Engineers (ASM E) Boiler and Pressure Vessel Code (BPVC)

Section III Division 5, Rules for Construction of Nuclear Power Plant Components, High Temperature Reactors, 2017 Edition.

Guidance Documents:

NUREG-2245, Technical Review of the 2017 Edition of ASME Code,Section III, Division 5, High Temperature Reactors dated January 2023 (ADAMS Accession No. ML23030B636)

Regulatory Guide (RG) 1.87, Acceptability of ASME Code Section III, Division 5, High Temperature Reactors, Revision 2, dated January 2023 (ADAMS Accession No. ML22101A263)

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5.0 STAFF EVALUATION

5.1 Staff Evaluation Discussion

Kairos submitted this TR regarding the development of its safety-related reactor coolant boundary to support future licensing actions for reactors using the KP-FHR designs under 10 CFR Parts 50 or 52, including KP-FHR power reactors and non-power test reactors. The TR describes the qualification and testing methodology to be used for the metallic structural materials in safety-related components exposed to the high temperature reactor coolant salt (known as Flibe) environment of the KP-FHR designs. The Flibe properties are provided in the Kairos Power TR, Reactor Coolant for the Kairos Power Fluoride Salt-Cooled High Temperature Reactor, Revision 1 (ML20016A486), which was approved in an NRC staff SE dated July 16, 2020 (ML20139A224).

As stated in Section 5.1 of the TR, the sponsor requested NRC staff to review and approve the qualification requirements for environmental effects of Flibe on the metallic structural materials provided in Section 4 of the TR, which the applicant has proposed will partially satisfy PDC 14 and PDC 31. The qualification requirements provided in Section 4 of the TR are for environmental effects of Flibe on the metallic structural materials, which are in addition to the qualification requirements for mechanical properties of 316H austenitic stainless steel and ER16-8-2 stainless steel weld filler metal required by ASME Code,Section III, Division 5. The applicant stated that a description of how the remaining portions of these PDC are satisfied will be provided in safety analysis reports submitted with license applications for the KP-FHR designs. The applicant stated that these material qualification test results will be used as a basis in future licensing actions to address potential materials reliability and environmental compatibility issues via design, operation, and inspection.

The results of the planned tests and analyses, along with a description of the design, operation, inspection, and surveillance programs to manage the materials performance, will be provided in future license applications. The remainder of the TR was not evaluated by the NRC staff and was only reviewed as technical background and to identify any potential impacts on the portions of the TR for which Kairos requests approval. Therefore, KP-FHR designs referencing this TR may only use this TR for purposes related to the information on 316H and ER16-8-2 material found in Section 4 of the TR, subject to the specific Limitations and Conditions found in Section 6.0 of the NRC staff SE below. All other information related to 316H and ER16-8-2 material will be evaluated in separate documents and licensi ng actions (see Limitation and Condition 1).

As stated in Sections 1.1.3.2 and 5.1 of the TR, the reactor vessel is (( )) safety-related component exposed to Flibe that is required to keep the fuel covered in Flibe during all normal operations and postulated events. The environmental effects qualification testing in this TR was based on the environment that the reactor vessel would experience. Therefore, the environmental effects qualification testing for the KP-FHR designs in this TR can only be used for other components with environments that are bounded by the environment the reactor vessel would experience and referenced in this TR. For example, other components that would have Flibe on one side of the metallic material and another salt on the other side of the metallic material, or would be exposed to higher irradiation levels than those specified in the TR, or be subject to conditions otherwise not addressed in the TR would not be bounded by this TR (see Limitation and Condition 2.)

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The metallic structura l mate rials proposed for the KP-FHR designs a re 316H austenitic stainless steel and the associated ER16-8-2 stainless steel w eld filler metal w hich a re qual ified fo r use in ASME Code, Section Ill, D iv ision 5, fo r h igh temperature reacto rs. The NRC staff notes that 316H and ER 16-8-2 are mater ials that can be used in high temperature reactors since these mater ia ls are qualified mater ials listed in ASME Code, Section Ill, Division 5. ASME Code,

Section Ill, D ivision 5, prov ides min imum quality requ irements for the mater ia ls to ensu re the use of the mater ials w ill resu lt in an extremely low probab ility of abnormal leakage, rapidly propagating fa ilure, or g ross rupture, w hich partially satisfies PDC 14 and PDC 31. The NRC staff has endo rsed the use of ASME Code, Section Ill, D iv ision 5 as per NUREG-2245,

'Techn ical Review of the 2017 Edition of ASME Code, Section Ill, D iv ision 5, "H igh Tempe ra tu re Reactors ", ( ML23030B636 ) and Regu latory Gu ide 1.87, "Acceptability of ASME Section Ill, D ivision 5, High Tempe ratu re Reactors, " ( ML22101A263 ).

Ho w eve r, ASME Code, Section Ill, Div ision 5, specifies ER16-8-2 w eld fille r metal as a qualified mater ia l for use up to 650 °C w hile 316H is qualified for use up to 816 °C. Current ly, testing is be in erformed in accordance w ith the ASME Code to extend the ua lification of ER 16-8-2 up to [ 11 w h ich inc ludes the extens ion o e ma ena s s ress rup ure ac ors o e Ig er empera ure. Th is testing is addressed be lo w in further detail.

Although ASME Code, Section Ill, Div ision 5, contains stress rupture values up to [-11. the staff endorsement in RG 1.87 imposes a limitation to not endorse all the stress rupt~ues found in Tab le HBB-I -14.6B, " Expec ted Minimum Stress-to-Rupture Va lues, 1,000 psi ( MPa ),

Type 316 SS." The NRC staff limitation prov ides tab les to sho w acceptable use of the stress ru tu re data based on the amount of time at a specified tern eratu re

. Ho w ever, because Kairos stated that

e s a in s Is o e accep a e ecause e Ime a e spec I Ie emperature, fo r both normal operations and postu lated acc idents, falls w ithin the NRC staff endorsed ranges found in Tab le 2 of Regulatory Gu ide 1.87 for 316H. If the time and temperature for both normal operat ions and postu lated accident conditions change for the KP FH R designs, they must still be bounded by the NRC staff-endorsed ranges found in Tab le 2 of Regulatory Guide 1.87 for 316H, or an adequate justification must be prov ided for NRC staff review and app roval as to w hy the values outs ide of the endorsed ranges are acceptab le. (see Lim itation and Condition 3.)

Since ER16-8-2 is not cu rrently qua lified to the higher tempe ratu re necessary to support acc ident scenar ios of the KP-FH R desiiiii ns, the N RC staff imposes a cond ition that ER 16-8-2 must be qua lified to a tempe ratu re of [ 11 in accordance w ith the requirements of ASME Code, Section Ill, Div ision 5, tha oun s e postulated accident cond itions. The qualification must a lso be approved by the NRC staff (see Limitation and Cond ition 4 ).

5.1.1 Design of the KP-FHR

Section 1. 1 of the TR provides an overv iew of the key design features of the KP-FHR designs.

The applicant stated that these features are not expected to change du ring the development of the KP-FHR designs. The applicant a lso stated that these featu res provide the bas is for the safety review of the TR and that if fundamental changes occur to the key design featu res, or ne w or revised regu lations a re issued, these changes w ou ld be reconci led and addressed in

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future su bm ittals. Because the TR is reques ting app roval of certa in cha racter istics of the reacto r coolant boundary w ithou t the fu ll scope of kno w ledge of detailed system sp ecifications, there may be instances w here the design features, as o u tlined in the TR, change b etween su b mittal of this TR and a future licensing act ion. Acco rding ly, the NRC staff added a cond ition and limita tion to the TR con tingen t on the design features provided in Sec tion 1 of the TR (see Lim itation and Condition 5 ).

5.1.2 Env ironment to be Tested

The en vironmen ts fo r both the non - po w e r (test) reactor and the comme rc ial po w er reactor a re spec ified in Tab le 1 o f the TR and are similar except tha t the non-po w er reacto r lifetime is 5 yea rs, as opposed to [1111111111111)) for the commercia l p o w e r reac to r. The opera ting envi ronment pa rameters for the KP-~ igns concern ing envi ronmenta l degradation inc lude the fo llo w ing :

• F libe salt temperatures of 550°C-650 °C

• An intermediate sa lt coo lant loop for the commerc ia l reactor

• A Primary Hea t Transport Sys tem that rejects heat to the air in lieu of an intermediate coo lan t loo p for the non - po w er test reacto r

• Non - po w e r test reactor lifetime of 5 yea rs ( 1 yea r comm ission ing and 4 yea rs ope ra tion )

and commer cia l p o w e r reac to r life time of [-))

• "Near - atmos p heric " p rimary coolan t p ressu res

• End of life irrad ia tion of less than 0.1 d isplacement pe r a toms (d p a )

These a re key ope rating environment parame ters necessary to dev elop the qua lifica tion testing of 3 16H and ER 16 2 for spec ific env ironmental deg radation mechan isms. Therefore, the NRC staff is imposing a limita tion and cond ition that KP - F H R des igns referencing this TR must have the key opera ting environment pa rameters desc ribed above and, if changed, could necessita te the modifica tion o f, o r add ition to, the testing program. (see Limitation and Cond ition 6 ).

Tab le 1 1 of the TR p rov ides the sp ecific degrada tion mechan ics o f 316H and E R 16 2 for the o p era ting en vironment in the KP-F HR designs w ith the asso ciated testing to dete rmine the effects the ope ra ting envi ronment has on these mater ia ls. The NRC staff finds tha t en vironmen ta l effec ts testing at the norma l operating tempe ra tu res to valida te the degrada tion of 3 16H and E R 16 2 mater ial is acceptab le since it dup licates the envi ronment the ma te rial w ou ld ex e rience dur ing ope ration. A lso, the addi tional testing u sing h ighe r test tempe ratu res

[ )) w ill allow the applicant to develop environmen ta l de rada tion rates that ma b e d ur in ostu la ted accident scenar ios

u re I

ure OS

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in Sec tion 4 of the TR should b e increased to [-

11 (see Limita tion and Condi tion 7 ).

As stated in Sec tion 4.2.3 of the TR, mos t of the testing w ill be cond u cted in " Nomina l Flibe",

i.e., Flibe w hich has b een pur ified to m in imize w a te r and o ther ox id izing con taminants bu t not w ith excess bery llium metal to invoke redox contro l (i.e., Redox Con trolled Flibe ). The NRC staff finds that ma te rial testing in Nom inal Flibe w ill b ound the mater ials (316H and ER 16-8-2 ) in Redox Con trolled Flibe because Nom ina l Flibe has a higher ox id izing potential leading to increased de radation rates than in Redox Con trolled Flibe. Redox Cont rolled Flibe uses

[ 11 w hich reduces the concentration of tellurium and the oxidizing poten tia l in om m a I e, ereby lead ing to potent ially lo w er deg radation rates in Redox Con trolled Flib e.

T herefore, the NRC staff finds tha t the mater ia l testin in Redox Controlled Flibe can be u sed as a sensitivi stud to

e e in u ure Icense a pp,ca ions or e P-FHR 11 in Flibe has the poten tia l to form intermetallic phases in 316H an as no ed in Reference 7. Th is potential effect is addressed in Sec tion 5.1.3.3.2 of this S E w ith assoc iated Limitation and Condition 11, to determine the effects of [11111111111111111 on the mechan ica l properties o f 316H and asso ciated w e ld filler met~

Sec tion 4.2.3.3 of the T R describes two poten tia l ac ciden t scena rios fo r the comme rcial po w e r reacto r ( i.e., intermed iate sa lt ingress fo r [--))) and a ir ingress fo r [--11 into the Flibe salt) tha t w ould produce a s ec ific co~n o f these im u rities th~ect the safe - related com onents.

e re o re, a es an o e T R,

as escn e in ec I0n..., prov, e e propose impu rity testing fo r both sa lt and air tha t w ill cove r accident scena rios postulated in the trans ient safety ana lyses, and or igina lly defined in the mater ia ls Phenomena Iden tification and Ranking Ta ble ( P IRT ) review. In addition, the ingress of air impu rities is also acco u nted for and tested in comb ination w ith the in te rmediate sa lt from the intermedia te loop fo r the po w e r reactor. The NRC sta ff finds this ap proach acceptab le for deve lo ping the effect on co rros ion ra tes that b oth air and the intermed ia te sa lt may have on 3 16H and ER 16 2 because it w ill bound the ac ciden t conditions fo r the po w er reactor. T he NRC staff also finds tha t performin co rros ion testing of 3 16H and ER 16 2 in Nom inal Flibe w ith air (as an impur ity) for u p to [ 11 prov ides a reasona b le method o f developing corrosion rates in Nomina l Flibe w ith Im pun Ies for the non-po w er test reactor. The NRC staff also notes that the deta ils of the im pu rity testing (e.g., the concent ra tion o f contam inant ) have not been determined, as stated in Ta b le 13 o f the TR. The refo re, the specific condi tions of the impur ities in Nomina l Flibe, including contaminant chemistry, u sed in the impu rity effects testing on 3 16H and ER 16-8-2 shall bound the acc ident scena rios postu la ted in the transien t ana lyses doc u mented in the safety analys is reports fo r the KP-FHR designs (see Limitation and Cond ition 8 ).

5.1.3 Deg radation Mechan isms

T he T R prov ides the necessary mater ia l testing to dete rm ine the rate of deg radation of 3 16H and ER 16 2 in the environmen t of the KP-FHR designs using Flibe. T he test results w ill be used to confirm tha t safety-related reactor coolan t b oundary mater ia l unde r ope ra ting and

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postula ted accident cond itions have an ext remely low probabil ity of abno rma l leakage, ra p idly p ropagat ing failure, or g ross ru pture, w h ich partially satisfies the c riteria in PDC 14 and 31. The mater ia l testing of 316H and ER16 2 in F libe w ill be conducted for the follow ing deg rada tion mechan isms :

• Co rros ion ( includ ing genera l co rros ion, crevice cor rosion, thermal ag ing, e rosion /w ea r and co ld leg occlus ion )

• Env ironmentally ass isted c racking (includ ing stress corrosion c racking, envi ronmen ta l creep, and co rros ion fa tigue )

• Effects on meta llurgical pro p erties ( includ ing stress relaxation cracking, phase fo rmation embr ittlement, and thermal cyc ling )

• Irrad iation effects (in cluding irrad iation -affected cor rosion, irrad iation-assisted stress co rros ion cracking, and irradiation-induced emb rittlement )

References 11 and 12 to the SE descr ibe var ious deg rada tion mechan isms, w hether they occu r in molten salt envi ronments, and w he re additiona l information may be needed. These references ident ify co rros ion, env ironmentally ass isted c rac king, and the effects of irradiation on mater ia ls as sub jects w he re kno w ledge ga p s may exist and require additiona l study. Reference 1 1 iden tifies tha t more da ta is needed fo r co rrosion in mol ten sa lt includ ing the effects o f imp urities and redox contro l on co rros ion ra tes, and tha t there is a kno w ledge gap fo r env ironmen ta lly assisted c racking in molten salts. This reference also notes that irrad ia tion may affect degradation of mater ia l in mol ten salts, but that little da ta is cur ren tly ava ilable.

Reference 12 iden tifies the poten tia l fo r fo rma tion of intermeta llic phases and the co rrespond ing reduc tion in mater ia l strength. The refore, the NRC staff finds tha t the above env ironmenta l degradat ion mechan isms a re pertinent to 316H and E R 16-8-2 in F libe and are consistent w ith information needs identified in current ly ava ila b le research da ta and testing descr ibed above and in the TR.

The refore, the NRC staff finds tha t the TR can be used in fu tu re licens ing ac tions for the a b ove degradat ion mechanisms desc ribed in Section 4 of the TR fo r the KP-FHR des igns to partially satisfy PDCs 14 and 3 1, su b ject to the Lim itations and Conditions found in Section 6.0 of the NRC staffs SE. The specific eva luation fo r the testing o f each deg radation mechan ism is p rovided b e lo w. The NRC staff notes that add itional info rmation and research on d ifferen t degradat ion mechanisms may become ava ila b le in the fu tu re. These d iffe ren t degrada tion mechan isms w ou ld requ ire add itiona l testing and w ou ld b e eva luated in fu tu re licens ing ac tions.

5.1.3.1 Co rros ion

Sec tion 4.2.3 o f the TR prov ides an overv iew of the p roposed co rros ion testing tha t w ill be used to deve lo p quantita tive co rros ion models fo r 316H stainless steel in a F libe env ironment. The NRC staff did not ma ke a finding w ith rega rds to the ove rview of the p roposed cor ros ion testing in Sec tion 4.2.3.

5.1.3.1.1 Co rros ion Tes t Sys tems

Sec tion 4.2.3.1 of the TR descr ibes the s stems that w e re deve lo Ka iros stated tha t the

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)) The refo re, test systems tha t incorpo ra te these ea ures are accep a e ecause ey ensure that the degradation phenomena described in Sec tion 4.2.3.3, " Co rros ion Testing," of the TR can be accoun ted fo r.

5.1.3.1.2 Com p os itional Analysis and E lec trochemical Potentia l ( ECP )

The NRC staff evaluated the proposed use o f com positiona l analysis to mon itor the redox conditions of Flibe and finds it acce ptab le b ecause it w ill quantify the impac t that the Flibe composi tion has on the co rros ion rates of the 3 16H and ER 16-8 -2 mater ia ls. An app licant referencing this TR fo r KP - FHR designs w ill need to demons tra te the Nomina l F libe composition fo r the coo lant is cons istent w ith the Nomina l Flibe compos ition (s) used in this qualification test prog ram (see Lim itation and Cond ition 9 ). Add itionally, the NRC staff finds the proposed use o f ECP mon itor ing dur ing testing accep ta b le because it w ill allow Kairos to measure the ingress o f ox id izing impu rities into the Flib e. The NRC staff a lso notes that the use of ECP du rin testin is acce tab le b ecause Kairos w ill a lso

5.1.3.1.3 Co rros ion Tes ting {Genera l Corrosion, Crevice Corrosion, E rosion/Wea r, Thermal Ag ing and Co ld Leg Oc clusion}

Sec tion 4.2.3.3 of the TR desc ribes the proposed co rros ion testing fo r 3 16H and ER1 6 2 ex posed to Flibe. The proposed testing w ill u se cou p ons of these ma terials in cond itions desc ribed in Ta b les 12 and 13 of the TR. Tests w ill be performed unde r differen t conditions and w ill also in clude tests in off-nom inal cond itions to assess the impacts o f spe cific corros ion

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degradation mechanisms. This includes tests to determine the effects of temperature, microstructure, salt composition, geometry, erosion-corrosion, thermal aging, graphite contact, and difference in solubility of corrosion products on the corrosion rate of 316H and ER16-8-2.

The NRC staff evaluated the planned corrosion testing for the KP-FHR designs that is summarized in Section 4.2.3.3, and Tables 12 and 13 of the TR. The staff also evaluated the proposed method to determine corrosion kinetics and the steady state corrosion rate, which are described in Section 4.2.3.3 and Appendix C of the TR. The NRC staff finds the proposed corrosion testing acceptable because these test s will determine the impact of temperature, microstructure, salt composition, geometry, erosion-corrosion, thermal aging, presence of graphite, redox control, and difference in corrosion product solubility (i.e., cold leg occlusion) on the corrosion rates and corrosion kinetics of 316H and ER16-8-2. In addition, these tests are acceptable because they are consistent with the expected corrosion mechanisms for 316H and ER16-8-2 in a molten salt environment (Raim an 2021) and a portion of the tests will be conducted with flowing Flibe, which is necessary to simulate the flowing salt in a reactor.

The NRC staff finds the tests to determine the effect of temperature on corrosion rates acceptable because corrosion is evaluated over a range of temperatures consistent with the operating temperatures of the KP-FHR designs including bounding postulated accident conditions which satisfies PDCs 14 and 31, in part. In addition, the NRC staff finds the test durations will provide sufficient data to determine corrosion kinetics.

The NRC staff also finds the tests to evaluate the microstructural effects on corrosion rates acceptable because, as described in Table 12 of the TR, these include tests to examine effects of (( )) which are known to increase corrosion rates.

The NRC staff finds that the tests using both the Nominal Flibe composition, as well as those tests with a reducing agent added, are acceptable because these tests will determine the effects of the Flibe composition, including how oxidizing contaminants, as well as redox control, affect the corrosion rate. These tests will provide data necessary to determine design margins for corrosion, allowable levels of impurities in the salt, and the potential benefit from adding a redox control agent. An applicant referencing this TR must demonstrate that the salt compositions (with reducing agent additions and impurities from postulated accident scenarios) tested in this program bound any potential salt compositions for the KP-FHR designs (see Limitation and Condition 10).

With regard to occluded geometry effects on corrosion rates, the NRC staff finds the proposed tests acceptable because these tests will determine whether crevice corrosion is a concern for 316H and ER16-8-2 in Flibe, and the potential effect on the corrosion rate.

The NRC staff finds the proposed tests to determine the effect of erosion-corrosion acceptable because the tests will utilize graphite particulate to determine the effect of these particles on corrosion rates, as well as (( )). This is necessary because the KP-FHR designs will utilize graphite pebbles, as well as a graphite reflector, which will introduce graphite dust into the Flibe. Additionally, it is appropriate to determine the effect of graphite on corrosion because the presence of graphite can accelerate corrosion of 316H and ER16-8-2 when in the same system as the fluoride salt (Flibe).

The NRC staff finds the tests to determine the impact of cold leg occlusion acceptable because the proposed tests have a temperature differential between the hot and cold legs consistent with

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the KP-FHR designs. This temperature differential is necessary because corrosion products are more soluble in the hot leg and will precipitate out in the cold leg. This creates a concentration gradient that will accelerate corrosion as a function of the temperature differential between the hot and cold legs and will be simulated in the tests.

5.1.3.1.4 Corrosion Modeling

Section 4.2.3.3 of the TR stated that testing will be used to analyze the depth of Chromium loss over time to establish the corrosion kinetics and to determine the steady state corrosion rate.

The depth of Cr loss and other metallurgical changes will be analyzed using electron microscopy. Appendix C, "Data Analysis", of the TR stated that this will allow for more sensitive measurements than analyzing the weight change of the test coupons. This is because measuring weight change can be complicated due to factors such as carbon pickup or difficulties in removing dried salt from the coupons. Electron microscopy will instead allow Kairos to analyze coupon cross sections to assess corrosion and other compositional changes.

Appendix C of the TR also stated that baseline corrosion models will be developed and separate effects tests will assess key variables that may impact corrosion rates. Kairos also stated that it will perform statistical analysis on the data and will utilize prediction bands to ensure appropriate and conservative extrapolation to the KP-FHR operational times and temperatures. For test data of certain degradation mechanisms (e.g., stress corrosion cracking) that may not be amenable to statistical analysis, Kairos stated that testing will be performed to detect if the phenomenon occurs, and whether variables that impact stress corrosion cracking can be quantified in order to perform a statistical analysis on the data. In scenarios such as this, Kairos stated that other practices (e.g., periodic inspections) may be used to address such phenomena, if the test data is not amenable to performing a statistical analysis.

The NRC staff evaluated the proposed corrosion modelling by Kairos in order to determine if the proposed qualification program for the KP-FHR designs will be adequate to determine performance of 316H and ER16-8-2 when exposed to the molten Flibe reactor coolant. The staff finds it acceptable to model corrosion behavior as a function of Cr loss from the 316H and ER16-8-2 because Cr is the alloying element in 316H that is most thermodynamically favored to corrode (i.e., least noble) and therefore will likely corrode prior to other elements of 316H and ER16-8-2 (DeVan, 1962, Raiman 2021). The staff also finds it acceptable to analyze the corrosion data as described in Appendix C because statistical analysis of the data will provide reasonable assurance that significant contributors to corrosion can be identified and that uncertainties resulting from the test data can be conservatively incorporated into corrosion predictions. Additionally, the staff finds use of electron microscopy acceptable because this will allow Kairos to assess the depth of Cr loss as well as other compositional changes in the material to mitigate complicating factors from the corrosion tests such as carbon pickup or difficulty removing dried salt from the material. This will provide data that can be corroborated against the observations from the electron microscopy. Use of electron microscopy is also acceptable because, as stated in Section 4.2.3.3 of the TR, weight change for each corrosion coupon will also be measured. The staff finds it acceptable to perform separate effects testing, in addition to baseline corrosion testing, because it will allow different variables to be assessed for their impacts on the corrosion rate. The staff finds it acceptable to perform some tests primarily to detect whether a specific phenomenon occurs, if the test data of a degradation mechanism is not amenable to statistical analysis, because after assessing whether a phenomenon occurs, it can be quantified and mitigated via multiple measures (e.g.,

inspections).

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5.1.3.1.1.5 Effects o f Operating Cond itions on Corros ion

ach to assess and manage cor rosion performance is acceptab le.

5.1.3.2 Env ironmentally Assisted Crack ing

5.1.3.2.1 Stress Cor rosion C rack ing and Co rrosion Fat igue

Section 4.2.4 and Tab les 14 and 15 o f the TR provides the proposed mater ia l testing that w ill be performed to eva luate ho w the ope rating env ironment of the KP-FHR designs using Flibe affects the cor rosion fatigue and stress co rros ion c rac king rates o f 316 H and E R16 2.

Cu rrently, there is little mechan ica l testing in molten sa lts due to the d ifficulty o f conduct ing in situ mechanical testing in h igh ly reducing mo lten sa lt. There is a lso limited data of env ironmenta lly assisted c racking in stainless steels and nickel-based a lloys in mo lten salts.

The refore, in-situ mechanica l testing systems w ill be used to conduct slo w strain rate testing (SS RT ) for co rros ion fatigue and stress cor rosion c rackin. Section 4.2.4.1 of the TR states that the SSRT tests w ill be conducted at temperatures [ 11 at var ious strain rates as desc ribed in Table 14 o f the TR. The SS RT testing w, e con ucted in Nom inal Flibe and Red ox Cont rolled Flibe to assess if 3 16 H, ER 16 2, and the [11111111111111111 of 316H a re suscept ib le to env ironmentally ass isted crack ing in Flibe. Secti~ states that the SSRT testing w ill be conducted in accordance w ith Amer ican Society for Testing and Materia ls (ASTM ) ASTM G129 - 00, "Standa rd P ractice fo r S lo w Strain Rate Testing to Eva luate the Suscept ib ility of Metallic Mate rials to E nv ironmentally Assisted C rack ing, " 200 0 Ed ition.

In add ition to the SSRT testing, Section 4.2.4.2 of the TR states that fracture mechanics-based testing of pre-c rac ked compact tension type samp les w ill be used to evaluate fatigue c rack g ro wth rates, stress co rros ion c rac king rates, and the fractu re behavior o f 316H and E R 16-8-2 in the Flibe env ironment at tempe ratu res [ 11 as desc ribed in Tab le 15 of the TR.

The test samp les w ill cons ist of [

lllln w ith a sha r flaw i.e., the

~ w e ld meta l, ea to be tested.

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1 1 ack w hich w I co rros ion testing) and stress corrosion cracking portion (a constant stress intens ity factor to initiate stress cor rosion under agg ressive testing conditions and then transition to condit ions that a re representative of the KP-FHR designs ).

The NRC staff finds that the samples to be tested are acceptable because they are representative o f the mater ia l and w e ldments to be used in the KP-F HR desi ns. In the use o f

ros ion omina l Flibe and Redox Con ro e I e acceptab l....... e in simu lating and determining the crack growth rates fo r

compact tens ion samp les acceptab le because these test methods a re w ell-established, are an accepted methodo logy, and have been effective in othe r env ironmenta lly ass isted crack ing testing used by licensees and the NRC staff.

5.1.3.2.2 Env ironmental Creep

Section 4.2.4.3 of the TR desc ribes the mate rial testing that w ill be susce tib ility of 3 16H and ER 16 2 to environmental creep in [

)) as described in Tab le 16 of the TR The NRC sta m s a pe ormmg creep es mg m oth No rma l Flibe us ing w elded base meta l samp les to include the base metal, w eld meta l and heat affected zone of the base metal acceptab le since it simu lates the mater ia l to be used and the environment o f the KP-FHR desi ns. In addition, the NRC staff finds it accep table that [ )) are no t requ ired to be performed un less sign ifican egra a I0n Is no e compare o creep tests performed in a ir, such as failure times outside of the 90% confidence in terva l from c reep tests performed in air, o r a change in fracture mode because the c reep tests in a ir w ou ld bound the results in Nomina l Flibe in the tempe ratu re range o f the KP-FHR designs. The NRC staff notes that the creep tests in Nom ina l Flibe a re to determine if the Flibe con tribu tes add itional deg radation beyond those determine from the c reep tests performed in a ir. If the testing determ ines Flibe has an additiona l effect on deg radation, additiona l testing w ould be requ ired to quantify any inc rease in deg radation contributing to F libe, and the test results w ou ld be review ed by the NRC staff in future licensing submitta ls by the applican t.

5.1.3.3 Metallurgica l Effects

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5.1.3.3.1 Stress Relaxat ion Cracking

Section 4.2.5 of the TR states that stress relaxation crack ing in 31 GH w ill be addressed by using test resu lts conducted in air up to temperatures of [-11 as discussed in Section 3.2 of the TR, and by conduct ing future ana lys is and design r~en ts of the KP-FHR designs, such as w eld designs in Figu re 23 of the TR, and specific w eld processes and parameters to m in imize stress relaxation c rack ing as detailed in Section 3.3.1 of the TR to reduce the triax ial stresses.

Section 3.3.2 of the TR rov ides

, w hile the heat affect one of 316H base metal w ith

• t... riaxial str * *

• 11.

The N RC staff finds testing in air acceptable because these test resu lts w ou ld be valid for 31 GH in Flibe fo r the KP-FHR designs since triaxial stresses are the major contr ibutor to stress relaxat ion c racking. In addition, the NRC staff finds that compa ring the susceptibility of 31 GH to that of 347 as discussed in Section 3.3.1 of the TR w ould allow a determination of the bounding triaxial stresses that could cause stress relaxation cracking in 316H. The NRC staff also finds the stress relaxation testing for the KP-FHR comme rcial o w er reactor and the non-o w er test reac tor in Table 1 O of the TR acce table because the

316H. The NRC staff notes that in Section 3.3.2 * *

• the ca abili to mode l w e ld residual stresses using [

11 to better asse idual f 31 GH fo r the KP-FHR designs. The results of these analysis can be used in future licensing actions to add ress stress relaxation c racking of 316H in the KP FHR designs.

5.1.3.3.2 Phase Format ion Embr ittlement

Section 4.2.5 of the TR d iscusses ho w the qualification prog ram addresses phase fo rmat ion embr ittlement, and deg radat ion from therma l cyc ling or thermal g rad ients. Kairos states that phase fo rmation embr ittlement may occur w hen 31 GH and ER16-8-2 picks up an e lement during its ex osure to Flibe and forms a deleterious second hase. To address this, Kairos ro osed to

The NRC staff review ed the proposed method to address * * *

• NRC staff finds it acce tab le because Ka iros w ill

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5 w hich states that the results of the qualification testing are on ly app licab le to the KP-FHR desi ns that is bound b the test conditions. In this case, a desi n that utilizes

, d to perform testing to qua ntify the effects on the mechanical properties of 3 16H and associated w e ld filler meta l ER 16-8-2 ( see Limitation and Condition 11 ).

5.1.3.3.3 The rmal Cycling /Stripping

Tab le 11 of the TR states that therma l cycl ing [

)]. In addition, Section 4.2.5 of the TR s a es a egra a,on o an - -

y g thermal transients cou ld lead to high stresses resulting in therma l fatigue degradat ion.

-Ka iros w ill add ress the thermal cycling by conducting analysis to refine the design and ope ration of the KP-FHR designs to m itigate large therma l g rad ients. The NRC staff finds it acceptab le that future ana lysis, in lieu o f testing, w ill be used to mitigate therma l cycling because the therma l gradients w ill be minim ized through the use o f approp riate design and operating conditions of the KP-FHR (po w e r and non - po w er test reacto r), as informed by the ana lys is.

Ho w eve r, since the design has not been fina lized and no testing w ill be conducted as part of this mater ia l qua lification program, the NRC staff is imposing a limitation and condit ion that an applicant implementing this TR w ill add ress therma l cycl ing / stripping in futu re licensing submitta ls by min im izing the thermal gradients v ia app rop riate design and ope rating cond itions of KP-FHR designs based on analysis (see Limitation and Cond ition 12).

5.1.3.4 Irrad iation Effects

5.1.3.4.1 Irrad iation-Induced Embr ittlement

Section 4.2.6.1 of the TR states that ex isting data indicates that tens ile properties and fracture toughness o f austen itic stainless steels, w hen tested at h igh strain rates and tempe ratu res from 550 °C to 650 °C, a re relatively unaffected by irrad iation levels <0.1 d isplacement pe r atoms (dpa ) w ith a helium content of 10 atom ic parts per m illion (appm ) in current light w ater reacto r env ironments. Ho w eve r, at lo w strain rates, data sho w s irrad iation-induced emb rittlement can affect mater ia l properties such as tensile strength and ductility and creep life due to the generation of he lium. The app licant stated in Section 4.2.6.1 of the TR that existing data w ill be used to deve lop deg radat ion facto rs, but that it w ill conduct irradiation tests on E R16 2, 316H,

and the associated heat affected zone of 316H to quantify ma rg ins at irradiation levels for the non-po w er test reactor and the comme rcia l po w er reactor w hich w ill be prov ided in future licensing act ions. The NRC staff finds it acceptab le to conduct testing fo r irrad iation-induced embr ittlement on ER16-8-2, 316H, and the associated heat affected zone of 316H, because the testing w ill be rep resentative o f the environment in the KP-FHR designs and this information w ill be subm itted in futu re licensing actions. NRC staff is imposing a lim itation and condition that the test environment shall bound the KP-F HR designs, including the expected irrad iation damage (dpa ) and helium content (see Limitation and Cond ition 13 ).

5.1.3.4.2 Irrad iation-Affected Co rros ion

Section 4.2.6.2 of the TR states that no immed iate mate rial testing o f 3 16H and E R16-8-2 fo r irrad iation effects on corros ion is proposed for the qualification of 316H and ER 16-8-2 because the reacto r vessel has a low irradiation dose level (< 0.1 dpa ) and existing data sho w s that

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irradiation may increase general corrosion rates but decrease intergranular corrosion rates.

However, the applicant will implement a mate rials surveillance system program for the non-power test reactor and (at least the first) commercial power reactor systems to monitor irradiation-affected corrosion. In addition, an inspection and monitoring program that will assess the wall thickness of the reactor vessel will also be implemented. The initial plans for these programs are provided in Appendix B of the TR. The applicant has not finalized plans for these programs and will provide the detailed programs in future licensing actions.

Since Appendix B of the TR is not a finalized program for assessing irradiation-affected corrosion, the NRC staff cannot provide a conc lusion on the proposed initial planned programs.

Notwithstanding, the NRC staff finds it accept able to implement a materials surveillance program that will be submitted as part of future license applications for the non-power test reactor and the commercial power reactor because this program could provide sufficient information that can be used in determining any affects irradiation has on the corrosion rate of 316H and ER16-8-2 in the environment of the KP-FHR designs. However, the NRC staff notes that the materials surveillance program should not be limited to only the first commercial power reactor, because there is limited data on the effects of irradiation on corrosion rates in Flibe on 316H and ER16-8-2. Therefore, the materials surveillance program should apply to both the non-power test reactor and the commercial power reactors. In addition, the NRC staff finds it acceptable to use an inspection and monitoring program to assess any changes in the wall thickness of the reactor vessel because the program should be capable of detecting wall thinning that could prevent the reactor vessel from performing its safety function. Therefore, the NRC staff is imposing a limitation and condition that the materials surveillance program and the inspection and monitoring program will be submitted in future license applications for NRC staff review and approval to verify that these programs are sufficient to address irradiation-affected corrosion of the reactor vessel. (See Limitation and Condition 14.)

5.1.3.4.3 Irradiation-Assisted Stress Corrosion Cracking (IASCC)

Section 4.2.6.3 of the TR states that IASCC is not expected to be a degradation mechanism in the KP-FHR design due to the low irradiation level (<0.1 dpa) and that radiolysis of Flibe is not expected because of the rapid recombination of ions in the molten Flibe state. In addition, the chemistry control system will have the capability to adjust the redox potential of the salt and to correct Flibe chemistry changes induced by transmutation. The applicant also states that the test program specified in Section 4.2.4 will determine if stress corrosion cracking is a credible degradation mechanism for the environment of the KP-FHR designs. Therefore, the applicant does not propose additional material testing of 316H and ER16-8-2 for irradiation effects on stress corrosion cracking. However, a materials surveillance program and the inspection and monitoring program, as discussed in Appendix B of the TR, will be implemented and submitted in future license applications to address concerns for IASCC.

The NRC staff finds it acceptable to implement a materials surveillance program that will be submitted in future license applications for the non-power test reactor and the commercial power reactor because this program could provide sufficient information that can be used in determining any effects irradiation has on the stress corrosion cracking rate of 316H and ER16-8-2 in the KP-FHR environment. As stated in Section 5.1.3.4.2 of this SE, the NRC staff notes that the materials surveillance program should not be limited to only the first commercial power reactor, because there is limited data on the effects of irradiation on stress corrosion cracking rates in Flibe on 316H and ER16-8-2. Consistent with the discussion in Section 5.1.3.4.2 of this SE, above, this warrants implementation of a materials surveillance program for all commercial

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po w er reac to rs us ing the KP-FHR design. In add ition, the NRC staff finds it acceptab le to use an inspection and mon itor ing prog ram to detect cracking of the reac to r vessel b ecause the program should be capab le of detecting c racking tha t w ould preven t the reactor vessel from performing its safety func tion. Therefore, the NRC staff is imposing a limitation and cond ition that the ma terials su rveillance prog ram and the inspec tion and mon itoring prog ram w ill be su b mitted in fu tu re license app lications fo r NRC staff review and app rova l to verify tha t these programs a re sufficient to address irradiation - affected stress cor ros ion c racking o f the reactor vessel (see Limitation and Cond ition 14 ).

5.1.4 Qual ity Assurance

Sec tion 1 of the TR states tha t the non - po w e r test reactor a pplication is imp lementing a qual ity assu rance program based on ANS I/ANS-15.8-1995, "Quality Assurance Prog ram Requiremen ts fo r Resea rch Reactors, " (ANSI /ANS-15.8 ), w hich is endorsed by NRC Regu latory Guide 2.5,

"Quality Assurance Prog ram Requi rements fo r Research and Test Reactors." The NRC staff finds it acceptab le to use ANS I/ANS-15.8-1995 fo r ma te rial testing that w ill only be used to su pport the non-po w er test reactor. The NRC staff notes that Rev ision 4 of the TR does not specify if mater ia l testing related to safety-related componen ts w ill be conducted unde r a prog ram that com plies w ith the requ iremen ts of 10 CFR 50 Append ix B, as stated in previous revisions of the TR. The qua lity and accuracy o f mater ia l testing resu lts that cou ld be used fo r the commercia l po w e r reac to r must be confirmed w hen used to address potential mater ials reliability and env ironmental com patibility of safety-re la ted com ponents. Therefore, the NRC staff is impos ing a lim itation and condition that mate ria l testing w ill be conducted unde r a qual ity assu rance program tha t complies w ith the requ irements of 10 C FR 50 A ppend ix B to con firm the qua lity o f the da ta ob ta ined du ring the mater ia l testing that w ill be used fo r the commercia l po w er reac to r (see Limitation and Cond ition 15).

5.2 Eva luation Summary

The N RC staff finds that the mater ia l qualification methodo logy for 3 16H and ER 16-8-2 mater ia ls in Section 4 o f the TR sa tisfy, in part, the PDCs 14 and 3 1 fo r the KP-FHR designs and is acce ptab le, sub jec t to the Limitations and Condi tions found in Section 6.0 of the NRC staff' s S E belo w. The NRC staff finds that testing a t the no rma l o pe rating and postulated acc iden t temperatures, and in both Nom inal Flibe and Redox Cont rolled Flibe, to va lidate the degradat ion o f 316H and E R16 2 mater ia l, is accepta ble since the testing dup licates the o perating env ironment that the ma terial w ill expe rience in the KP-F HR designs. The NRC staff a lso finds it acceptab le tha t the mater ial test sam les w ill include no t onl the 3 16H base meta l and asso ciated ER 16-8-2 w e ld me ta l, bu t the

e sa aso in s a e re Is reasona e assurance a e eg ra a I0n mec anisms to be tested as desc ribed in in Sec tion 4 of the TR incl u de the a ppro priate environmental deg radation mechan isms fo r the KP-FHR designs based on the cur rent research and testing information provided in the TR and in References 11 and 12 o f this SE. These refe rences d isc u ss top ics such as cor ros ion, environmentally ass isted c rac king, and the effects of irrad iation on mater ials, and their ap plica b ility in mo lten sa lt environments.

The staff has reasonab le assurance the qua lification program meets the requ iremen ts listed in Sec tion 4. 1 descr ibed above, as they relate to the q u a lification o f 316H and E R16 2 in the

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Flibe environment, because the TR describes the use of generally accepted engineering standards, unique safety features, novel design features, and the relation of facility design to the PDC.

6.0 LIMITATIONS AND CONDITIONS

An applicant may reference the TR only if the applicant demonstrates compliance with the following limitations and conditions:

1. (Section 1.0) As stated by Kairos in the TR, NRC staff review and approval of only Section 4 of the TR was requested. Therefore, KP-FHR designs referencing this TR may only use this TR for purposes related to the information on 316H and ER16-8-2 material found in Section 4 of the TR, subject to the specific limitations and conditions found in the NRC staff SE below. All other information related to the 316H and ER16 2 material will be evaluated in separate documents and licensing actions.

2. (Sections 1.1.3.2 and 5.1) The environmental effects qualification testing for the KP-FHR designs in this TR can only be used for other components with environments that are bounded by the environment the reactor vessel would experience and are used in this TR. For example, other components that would have Flibe on one side of the metallic material and another salt on the other side of the metallic material, or higher irradiation levels than those specified in the TR, etc. would not be bounded by this TR.

3. If the time and temperature for both normal operations and postulated accident conditions change for the KP-FHR designs, th ey must still be bounded by the NRC staff-endorsed ranges found in Table 2 of Regulatory Guide 1.87 for 316H, or an adequate justification must be provided for NRC staff review and approval for why the values outside of the endorsed ranges are acceptable.

4. (Section 4.2.1) ER16-8-2 material must be qualified to a temperature of ((

)) in accordance with the requirements of ASME Code,Section III, Division 5, and for a time that bounds the postulated accident conditions and be approved by the NRC staff.

5. (Section 1.1) Because there is information that has not yet been developed and/or reviewed as part of this TR, KP-FHR designs referencing this TR must provide information that completely and accurately describes the design of the reactor coolant boundary (and associated systems) and any associated functions it is credited to perform for NRC staff review and approval. As stated in the TR, if key design features of the KP-FHR designs change, or if new or revised regulations are issued that impact descriptions and conclusions in this TR, these changes would be reconciled and addressed in future license application submittals. Due to the potential for design changes and new or revised regulations, KP-FHR designs referencing this TR must demonstrate that all regulatory and safety requirements related to the characteristics of the metallic materials are met when considering the final design of the KP-FHR.

6. (Section 4.1) As presented in the TR, there are key design parameters without which the proposed reactor coolant boundary design and associated properties may not be supported. Therefore, KP-FHR designs referencing this TR must have the following:

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• Flibe Salt temperatures of 550 °C-6S0°C

• An intermediate sa lt coolant loop for the commerc ia l reactor

• A Primary Heat Transport Sy stem that rejects heat to the air in lieu of an intermediate coolant loop fo r the non-po w er test reacto r

• Non-po w e r test reacto r lifetime of a maximum of 5 yea rs ( 1 year comm issioning~ ope ration ) and commercia l po w er reacto r lifetime of a max imum of [-11

• "Near-atmospher ic " primary coolant pressures

• End of life irradiation of less than 0.1 dpa

These key design parameters of the KP-FHR designs, if changed, cou ld necess itate the modification of, o r add ition to, the testing program.

7.

8. (Sec tio n 4. 2.3.3 a nd Ta bl e 13) The impur ity effects testing on 316H and ER16-8-2 must inc lude the potential loss of F libe chem istry contro l from both a ir ingress and intermediate sa lt loop ingress based on the sa fety ana lys is reports. An app licant referencing this TR must demonst rate that any potential impu rity ingress (inc luding postu lated acc idents ) in the KP-F HR designs is bound by the testing performed as part of this TR.

9. (S ec tio n 4.2. 3.2, Ta bl es 13 a nd 14) An applicant referencing this TR must demonst rate that the Nomina l F libe sa lt composition used in the KP-F HR designs is cons istent w ith the Nom inal Flibe salt composition used in these tests including initial impur ities in the sa lt.

10. S ect ion 4.2. 3.2, T a bl es 13 a nd 14) An app licant referencing this top ica l report mus t demonst ra te tha t the sa lt compos itions (w ith reducing agen t additions and impurities from postu la ted acc ident scenar ios ) tested in this program bound any poten tia l sa lt composi tions for the KP-F HR reac to r designs.

11. (Sec tio n 4.2. 5) In o rder to add ress phase fo rmation embr ittlemen t for the KP-FHR desi ns an a licant mus t sho w that testing bounds poten tia l design conditions 11 and that if a secondary phase is detected unng es mg, e e ec son mec anica properties of 316H and ER 16 2 must be quan tified via testing and app roved by the NRC staff.

12. (Sec tio n 4.2. 5 a nd T a bl e 11 ) The ap plicant w ill assess therma l cycl ing / striping in future licensing subm ittals by m inimizing the therma l gradien ts v ia app ropr iate design and o perating cond itions of the KP-F HR designs b ased on ana lysis.

13. (S ec tio n 4.2. 6.1 ) Testing fo r irradiation - induced emb rittlemen t of ER16-8 - 2, 3 16H, and the asso ciated heat affected zone of 3 16H must be performed that bounds the env ironmen t represen tative of the KP-F HR designs, inc luding the expected irrad ia tion

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damage (dpa) and helium content. The program describing this testing must be submitted in future license applications for NRC staff review and approval to verify this testing program is sufficient to address irradiation-induced embrittlement of the reactor vessel.

14. (Sections 4.2.6.2 and 4.2.6.3) As described in Sections 4.6.2.2 and 4.2.6.3 of the TR, a materials surveillance program and an inspection and monitoring program must be implemented for all non-power test reactors and commercial power reactors using KP-FHR designs to assess and monitor both irradiation-affected corrosion rates and irradiation-affected stress corrosion cracking rates of 316H and ER16-8-2 in the environment of KP-FHR designs. The materials surveillance program and the inspection and monitoring program must be submitted in future license applications for NRC staff review and approval to verify these programs are sufficient to address both irradiation-affected corrosion and irradiation-affected stress corrosion cracking of the reactor vessel.

15. (Section 1.0) Material testing for the commercial power reactor must be conducted under quality assurance program that meets the requirements of 10 CFR Part 50 Appendix B to confirm the quality of the data obtained during the material testing that will be used for the commercial power reactor.

7.0 CONCLUSION

Based on the evaluation above, the NRC staff concludes that Kairos has provided reasonable assurance that the information in Section 4 of the TR will satisfy, in part, KP-FHR PDCs 14 and 31 as described above, for the KP-FHR designs subject to the Limitations and Conditions in Section 6.0 of this SE. The NRC staff also concludes that the qualification program proposed by Kairos will satisfy, in part, the requirements of 10 CFR 50 and 52, as described in Section 4.1 above, with respect to contents of applications, subject to the limitations and conditions discussed above. The information provided in Section 4 of the TR establishes the material qualification methodology for environmental effects of Flibe on the 316H and ER16-8-2 structural materials to be used as a basis in future licensing actions to address potential materials reliability and environmental compatibility issues of the reactor vessel using the KP-FHR designs. The results of the planned tests, along with a description of the design, operation, inspection, and surveillance programs to manage the materials performance must be provided as part of future license application submittals.

8.0 REFERENCES

1. Kairos Power LLC letter No. KP-NRC-2006-004, dated June 30, 2020 (ADAMS Accession No. ML20182A800) submitting Kairos Power LLC, Metallic Materials Qualification for the Kairos Power Fluoride Salt-Cooled High-Temperature Reactor, KP-TR-013, Revision 0, June 30, 2020 (ADAMS Accession No. ML20182A800)

2. Email, Nuclear Regulatory Commission Richard Rivera to John Price, Preliminary Questions on Kairos Metallic Materials Qualification Topical Report, November 25, 2020 (ML20332A076).

3. Kairos Power LLC letter No. KP-NRC-2106-007, KP-FHR High Temperature Metallic Materials Topical Report, KPTR-013, Revision 1, June 30, 2021 (ML21181A386)

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submitting KP-FHR High-Temperature Metallic Materials Topical Report, KPTR-013, Revision 1, June 30, 2021 (ML21181A387)

4. Email, Nuclear Regulatory Commission Richard Rivera to Darrell Gardner and John Price, Preliminary Questions on Revision 1 of Kairos Metallic Materials Qualification Topical Report, October 13, 2021 (ML20332A076)

5. Kairos letter No. KP-NRC-2204-003, KP-FHR High Temperature Metallic Materials Topical Report, KPTR-013, Revision 2, dated April 26, 2022, (ML22116A247) submitting Metallic Materials Qualification for the Kairos Power Fluoride Salt-Cooled High-Temperature Reactor, KPTR-013, Revision 2, April 2022 (ML22116A249)

6. Kairos letter No. KP-NRC-2208-001, KP-FHR High Temperature Metallic Materials Topical Report, KPTR-013, Revision 3, dated August 19, 2022, (ML22231B222) submitting Metallic Materials Qualification for the Kairos Power Fluoride Salt-Cooled, High Temperature Reactor, KPTR-013, Revision 3, April 2022 (ML22231B224)

7. Kairos letter No. KP-NRC-2209-005, KP-FHR High Temperature Metallic Materials Topical Report, KP-TR-013, Revision 2, dated September 20, 2022, (ML22263A457) submitting Metallic Materials Qualification for the Kairos Power Fluoride Salt-Cooled, High Temperature Reactor, KP-TR-013, Revision 4, September 2022 (ML22263A459)

8. Kairos Power LLC, letter KP -NRC-1907- 006, P. Hastings, Vice President, Regulatory Affairs and Quality, to USNRC document control desk, re: Principal Design Criteria for the Kairos Power Fluoride Salt -Cooled, High Temperature Reactor (Revision 1),

July 31, 2019 (ADAMS Accession No. ML19212A756).

9. US NRC, NUREG-2245, Technical Review of the 2017 Edition of ASME Code,Section III, Division 5, High Temperature Reactors, dated January 2023 (ML23030B636).

10. US NRC, Regulatory Guide 1.87, Acceptability of ASME Section III, Division 5, High Temperature Reactors, Revision 2, dated January 2023 (ML22101A263).

11. Stephen S. Raiman, et. al., Oak Ridge National Laboratory, TLR-RES/DE/CIB-CMB-2021-03, Technical Assessment of Materials Compatibility in Molten Salt Reactors, March 2021 (ADAMS Accession No. ML21084A039).

12. J. R. Keiser, P. M. Singh. M.J Lance et. al., Interaction of Beryllium with 316H Stainless Steel in Molten Li2BeF4 (Flibe), Published in Journal of Nuclear Materials, Volume 565, July 2022.

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