ML23011A104

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

ML23011A104
Person / Time
Site:	99902069, Hermes  File:Kairos Power icon.png
Issue date:	01/11/2011
From:	Rivera R NRC/NRR/DANU/UAL1
To:
References
EPID L-2020-TOP-0050, KP-TR-013
Download: ML23011A104 (1)

Text

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UN ITE D STATES NUCLEAR REGULATORY COMMISS ION WA SHINGTON, D.C. 20555-0001

2022

DRAFT SAFET Y EVALUATION OF METALLIC MATERIAL QUALIFICATION FOR THE KAIROS POWER FLUORIDE SALT-C OOLED HIGH-TEMPERATURE REACTOR

( KP -TR- 0 13) KAIROS POWER, LL C EPID NO. 000 4 31 / 9990 206 9 / L-2 020-TOP- 0050

THIS NRC STAFF DRAFT SAFETY EVALUATION (SE) HAS BEEN PREPARED AND IS BEING RELEASED TO SUPPORT INTERACTIONS WITH THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS (ACRS). THIS DRAFT SE HAS NOT BEEN SUBJECT TO NRC MANAGEMENT AND LEGAL REVIEWS AND APPROVALS, AND ITS CONTENTS SHOULD NOT BE INTERPRETED AS OFFICIAL AGENCY POSITIONS.

1.0 SPONSOR INFORMATION

Sponsor: Ka iros Po w er, LLC ( Ka iros )

Address: 707 West To w er Ave.

Alameda, CA 94501

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

2.0 SUBMITTAL. CORR ESPONDENC E. AND CONTR IBUTORS

2.1. Submitta l Information

Rev ision 0 June 30, 2020 ML20182A799 KP-TR - 014, Rev ision 0 Rev ision 1 June 30, 2021 ML2 1181A385 KP-TR - 014, Rev ision 1 Rev ision 2 l\\pril 2, 2022 ML22116A246 KP-TR - 014, Rev ision 2 Rev ision 3 A.ugust19, 2022 ML22231 B22 1 KP-TR - 014, Rev ision 3 Rev ision 4 September 20, 2022 M L22263A456 KP-TR - 014, Rev ision 4

• Agenc yw ide Documents Access and Management System (ADAMS ) Accession No.

2.2. NRC Correspondence and Communicat ions

Communication Tvpe Date A.DAMS Access ion No.

Acceptance Review( s): September 3, 2020 M L20224A 172 Closed Meeting Notices: December 6, 2021 ML2 1336A400 Feb ruary 3, 2022 M L22032A336 Feb ruary 14, 2022 M L22032A336 Uulv 18, 2022 ML22196A385

~uaust10,2022 ML22214A13 1 September 12, 2022 ML22244A250

• ADAMS Access ion 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. Kairo s 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 reflects 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 (ASME) 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 August 2022 (ADAMS Accession No. ML21223A097 TBD)

Regulatory Guide (RG) 1.87, Acceptability of ASME Code Section III, Division 5, High Temperature Reactors, Revision 2, dated August 2022 (ADAMS Accession No. ML21091A276 TBD)

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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 aus tenitic 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 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 2 material will be evaluated in separate documents and licensing 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 3 16H austenitic stain less steel and the associated ER16-8-2 stain less steel w eld filler metal w h ich a re qualified fo r use in ASM E Code, Section Ill, D iv ision 5, fo r high temperatu re reacto rs. The NRC staff notes that 3 16H and ER 16-8-2 are mater ia ls that can be used in high temperature reactors since these mater ia ls are qua lified mate ria ls listed in ASME Code, Section Ill, D ivision 5. ASM E Code,

Section Ill, D iv ision 5, prov ides minimum qua lity requ irements for the mater ia ls to ensure the use of the mater ia ls w ill result in an extreme ly lo w probab ility of abnorma l leakage, rap id ly propagating fa ilure, or gross ruptu re, w hich partia lly satisfies PDC 14 and PDC 31. The NRC staff has endo rsed the use of ASME Code, Section Ill, Div ision 5 as per NUR EG-2245,

'Technical Rev iew of the 2 0 17 Ed ition of ASME Code, Section Ill, D iv ision 5, "H igh Temperature Reactors ", ( ML21223A097 TBD ) and Regu latory Guide 1.87, "Acceptab ility of ASME Section Ill, D ivision 5, High Temperatu re Reactors," ( ML21 091A276 TBD ).

Ho w eve r, ASME Code, Section Ill, Division 5, specifies ER 16-8-2 w e ld filler meta l as a qua lified mater ia l for use up to 650°C w hile 316H is qua lified fo r use up to 816°C. Cur rent ly, testing is bein erformed in acco rdance w ith the ASME Code to extend the ua lification o f ER 16 2 up to [ 11 w hich includes the extens ion o e ma en a s s ress rup ure ac o rs o e Ig er empe ra u re. This testing is addressed belo w in further deta il.

Although ASME Code, Section Ill, Div ision 5, contains stress rupture va lues up to [_11, the staff endorsement in RG 1.87 imposes a limitation to not endorse all the stress rupturevfflues found in Table HBB-I-14.6B, "Expected Min imum Stress-to-Ruptu re Va lues, 1,000 psi ( MPa ),

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

. Ho w ever, because Ka iros stated that

e s a in s Is o e accep a e ecause e,me a e spec,,e empe ratu re, fo r both no rma l ope rations and postu lated accidents, falls w ithin the NRC staff endorsed ranges found in Table 2 o f Regulatory Guide 1.87 fo r 316H. If the time and tempe ratu re fo r both no rma l ope rations and postu lated accident cond itions change fo r the KP FHR designs, they must still be bounded by the NRC staff-endorsed ranges found in Table 2 o f Regulato ry Gu ide 1.87 for 316H, o r an adequate justification must be prov ided fo r NRC staff review and app rova l as to w hy the va lues outside of the endo rsed ranges a re acceptab le. (see Limitation and Cond ition 3.)

Since ER16 2 is not cu rrently qual ified to the higher tempe ratu re necessary to support acc ident scenar ios o f the KP-FHR desiii!i ns, the NRC staff imposes a cond ition that ER16 2 must be qual ified to a temperatu re of [ 11 in acco rdance w ith the requirements of ASME Code, Section Ill, D iv ision 5, tha oun s e postulated accident cond itions. The qua lification must also be app roved by the NRC staff (see Limitation and Cond ition 4 ).

5.1. 1 Desig n of the KP-F HR

Section 1.1 o f the TR provides an ove rview of the key design features of the KP-F HR designs.

The app licant stated that these featu res are not expected to change du ring the deve lopment o f the KP-F HR designs. The applicant also stated that these featu res prov ide the bas is for the safety review of the TR and that if fundamenta l changes occur to the key design features, or ne w or revised regu lations a re issued, these changes w ou ld be reconc iled and add ressed in

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fu ture subm ittals. Beca u se the TR is reques ting app roval of certain cha racteristics of the reac to r coo lant boundary w ithou t the fu ll scope of kno w ledge of deta iled system specifications,

there may b e instances w here the design fea tures, as ou tlined in the TR, change be tween su b m ittal of this TR and a fu tu re licensing ac tion. Accord ingly, the NRC staff added a cond ition and limitation to the TR con tingent on the design featu res pro v ided in Section 1 of the TR (see Limitation and Cond ition 5 ).

5.1. 2 Env ironment to b e Tested

The environmen ts for b oth the non-po w er (test) reacto r and the comme rc ial po w e r reactor are specified in Tab le 1 of the TR and are sim ila r except that the non - po w er reac to r lifetime is 5 years, as opposed to [.... )] for the commer cia l po w er reactor. The o perating env ironmen t parameters for the KP-~ igns concern ing env ironmental degradation inc lude the fo llo w ing :

• Flibe salt tempera tures of 550 °C-650 °C

• An intermed ia te salt coo lant loo p for the commercia l reactor

• A Primary Heat Transport System tha t rejects heat to the air in lieu o f an intermed ia te coolan t loop fo r the non - po w er test reac to r

• Non - po w er test reactor lifetime of 5 yea rs ~omm issioning and 4 yea rs o peration )

and commercial po w e r reactor lifetime of [-)]

• " Nea r-atmosphe ric" primary coolan t pressu res

• End o f life irradiation of less than 0.1 displacement per a toms (dpa )

These are key opera ting env ironment pa rameters necessa ry to develop the qual ification testing of 316H and ER16 2 for specific env ironmenta l deg radat ion mechanisms. The refo re, the NRC staff is imposing a limitation and cond ition that KP-FHR designs referencing this TR must have the key o pe rating env ironment parameters desc ribed a b ove and, if changed, cou ld necess itate the mod ification of, or add ition to, the testing program. (see Limitation and Cond ition

6).

Tab le 11 of the TR provides the spec ific degradation mechanics o f 3 16H and ER 16 2 fo r the o perating environmen t in the KP-FHR designs w ith the associa ted testing to determine the effects the o pera ting environment has on these mate ria ls. The NRC staff finds that environmental effects testing at the no rma l o perating tem pe ratu res to va lidate the deg radation of 3 16H and ER 16-8-2 ma teria l is acce ptab le since it duplica tes the environmen t the mater ia l w ould ex erience du ring ope ra tion. Also, the additiona l testing us ing h ighe r test tempera tures

[ )] w ill a llo w the a pplicant to deve lo p env ironmenta l de radation rates tha t ma be aur in os tu lated acciden t scena rios

re I

ure OS

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in Section 4 of the TR shou ld be inc reased to [-

11 (see Limitation and Cond ition 7 ).

As stated in Section 4.2.3 of the TR, most of the testing w ill be conducted in "Nominal Flibe",

i.e., Flibe w hich has been pur ified to minim ize w ater and other ox id izing contaminants but not w ith excess beryllium meta l to invoke redox contro l ( i.e., Redox Controlled Flibe). The NRC staff finds that mater ia l testing in Nominal Flibe w ill bound the mate rials (3 16H and ER 16-8-2 ) in Redox Cont rolled Flibe because Nom ina l Flibe has a higher ox idizing potent ia l leading to increased de radation rates than in Redox Cont rolled Flibe. Redox Cont rolled Flibe uses

[ 11 w hich reduces the concentrat ion of tellurium and the oxidizing potential in omma I e, ereby leading to potent ially lo w e r deg radation rates in Redox Cont rolled Flibe.

Therefore, the NRC staff finds that the mater ia l testin in Redox Controlled Flibe can be used as a sensitiv i stud to

m u u re Icense app Ica ions o r e P-FHR 11 in Flibe has the potent ia l to fo rm intermetallic phases in 316H an as no ed in Reference 7. Th is potential effect is addressed in Section 5.1.3.3.2 of this SE w ith assoc iated Limitation and Cond ition 11, to determine the effects of [~11 on the mechan ical properties of 3 16H and associated w eld filler met~

Section 4.2.3.3 of the TR describes two potent ia l accident scenarios for the comme rcial po w e r reacto r ( i.e., intermediate sa lt ing ress for [..... ))) and air ing ress fo r [..... 11 into the Flibe salt) that w ould produce a s ecific co~n of these im urities th~ect the safe - re lated com onents.

e re ore, a es an o e TR,

as escn e m ec I0n..., prov I e e propose impu rity testing fo r both sa lt and a ir that w ill cover accident scena rios postulated in the trans ient safety analyses, and or ig inally defined in the mater ia ls Phenomena Ident ification and Rank ing Tab le ( PIRT ) review. In addition, the ing ress of a ir impur ities is also accounted for and tested in comb ination w ith the intermed iate sa lt from the intermed iate loop fo r the po w er reactor. The NRC staff finds this app roach acceptable for deve lop ing the effect on co rros ion rates that both a ir and the intermed iate sa lt may have on 316H and ER16 2 because it w ill bound the accident cond itions fo r the po w er reacto r. The NRC staff also finds that perform in - or ros ion testing of 316H and ER16-8-2 in Nomina l Flibe w ith a ir (as an impu rity) for up to [ )] prov ides a reasonable method o f develop ing cor ros ion rates in Nomina l Flibe w ith Impun Ies for the non-po w er test reactor. The NRC staff a lso notes that the deta ils of the impu rity testing (e.g., the concent ration of contam inant ) have not been determined, as stated in Tab le 13 of the TR. The refo re, the specific cond itions of the impurities in Nomina l F libe, including contam inant chemistry, used in the impu rity effects testing on 316H and ER16-8 - 2 shall bound the accident scena rios postu lated in the transient ana lyses documented in the safety analysis reports fo r the KP-FHR designs (see Limitation and Condition 8).

5.1.3 Deg radation Mechan isms

The TR p rov ides the necessary mater ia l testing to determ ine the rate of deg radation of 316H and ER 16-8-2 in the environment of the KP-FHR designs using Flibe. The test results w ill be used to confirm that sa fety-re lated reacto r coolant boundary mate rial under operat ing and

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postulated accident cond itions have an extreme ly low probab ility of abnorma l leakage, rapidly p ropagating fa ilure, or gross ruptu re, w hich partially satisfies the criteria in PDC 14 and 31. The mater ia l testing of 3 16H and ER 16-8-2 in Flibe w ill be conducted for the follow ing deg radation mechanisms:

• Corros ion ( includ ing genera l corrosion, c revice corros ion, therma l aging, eros ion /w ear and co ld leg occ lus ion )

• Env ironmentally ass isted crack ing ( including stress cor ros ion c racking, env ironmenta l creep, and corros ion fatigue)

• Effects on metallurgical properties ( including stress relaxat ion c rack ing, phase formation embrittlement, and therma l cycl ing )

• Irradiation effects (includ ing irradiation-a ffected cor ros ion, irradiation-assisted stress co rrosion cracking, and irrad iation-induced embr ittlement )

References 11 and 12 to the S E descr ibe var ious deg radation mechanisms, w hether they occu r in molten sa lt envi ronments, and w here add itional informat ion may be needed. These references identify co rros ion, env ironmentally ass isted crack ing, and the effects of irradiation on materials as sub jects w here kno w ledge gaps may ex ist and require add itional study. Re ference 1 1 identifies that more data is needed for co rros ion in mo lten sa lt inc luding the effects of impur ities and redox control on corrosion rates, and that there is a kno w ledge gap for envi ronmentally assisted c rack ing in molten sa lts. Th is reference a lso notes that irrad iation may affect degradat ion o f mater ia l in mo lten sa lts, but that little data is cur rent ly available.

Reference 12 identifies the potent ia l for formation of intermeta llic phases and the corresponding reduction in material strength. The refore, the NRC staff finds that the above environmenta l degradat ion mechan isms a re pertinent to 3 16H and ER 16 2 in Flibe and are cons istent w ith information needs ident ified in cur rent ly available research data and testing descr ibed above and in the TR

Therefore, the NRC staff finds that the TR can be used in future licens ing actions fo r the above degradat ion mechan isms desc ribed in Section 4 o f the TR for the KP-F HR designs to partially satisfy PDCs 14 and 31, subject to the Limitations and Cond itions found in Section 6.0 of the NRC staff's SE. The specific eva luat ion for the testing of each degradat ion mechan ism is p rov ided be lo w. The NRC staff notes that additiona l information and resea rch on different degradat ion mechan isms may become available in the future. These d ifferent degradat ion mechanisms w ou ld requ ire add itiona l testing and w ou ld be evaluated in future licensing act ions.

5.1.3. 1 Co rros ion

Section 4.2.3 of the TR prov ides an overv ie w of the proposed co rros ion testing that w ill be used to develop quantitat ive corros ion mode ls for 3 16H stain less steel in a F libe env ironment. The NRC staff d id not make a finding w ith rega rds to the overv iew of the p roposed co rros ion testing in Section 4.2.3.

5.1.3. 1. 1 Co rros ion Test Systems

Section 4.2.3.1 of the TR desc ribes the s stems that w ere deve lo Ka iros stated that the

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] Therefore, test systems that incorporate these ea ures are accep a e ecause ey ensure that the degradat ion phenomena described in Section 4.2.3.3, " Corrosion Testing," of the TR can be accounted for.

5.1.3.1.2 Compos itional Analysis and E lectrochemica l Potential ( ECP )

The NRC staff eva luated the proposed use of compositiona l ana lys is to mon itor the redox conditions of F libe and finds it acceptable because it w ill quantify the impact that the Flibe compos ition has on the corrosion rates of the 316H and ER 16-8-2 mater ia ls. An applicant referenc ing this TR for KP-FHR designs w ill need to demonstrate the Nomina l F libe compos ition for the coolant is cons istent w ith the Nom inal Flibe compos ition (s) used in this qual ification test program (see Limitation and Condition 9 ). Add itionally, the NRC staff finds the proposed use of ECP mon itoring dur ing testing acceptable because it w ill allow Kairos to measure the ingress of ox id izing impur ities into the Flibe. The NRC staff also notes that the use of ECP durin testing is acce table because Kairos w ill a lso

5.1.3.1.3 Corros ion Testing (General Corrosion. Crevice Corros ion. Eros ion /Wear. Therma l Ag ing and Co ld Leg Occlusion)

Section 4.2.3.3 of the TR describes the proposed corrosion testing for 316H and ER16-8-2 exposed to F libe. The proposed testing w ill use coupons of these mater ials in condit ions descr ibed in Tab les 12 and 13 of the TR. Tests w ill be performed under different cond itions and w ill also include tests in off-nomina l conditions to assess the impacts of specific corrosion

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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 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 tests 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 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 (Raiman 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 2 in Flibe, and the potential effect on the corrosion rate.

The NRC staff finds th e 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 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 th e 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 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 of Ope rating Cond itions on Corrosion

1 1 ach to assess and manage cor ros ion performance is acceptab le.

5.1.3.2 Env ironmentally Assisted Cracking

5.1.3.2.1 Stress Corrosion Cracking and Cor ros ion Fat igue

Section 4.2.4 and Tab les 14 and 15 o f the TR prov ides the proposed mater ia l testing that w ill be performed to evaluate ho w the ope rating environment o f the KP -FHR designs us ing F libe affects the cor ros ion fatigue and stress co rros ion crack ing rates of 316H and ER 16 2.

Cu rrently, there is little mechan ica l testing in mo lten sa lts due to the d ifficu lty of conducting in situ mechan ica l testing in high ly reduc ing molten sa lt. There is also lim ited data o f env ironmenta lly ass isted c rack ing in stain less steels and n ic ke l-based alloys in molten sa lts.

The refore, in-s itu mechan ica l testing systems w ill be used to conduct slow strain rate testing (SS RT ) fo r cor ros ion fatigue and stress cor ros ion c rack in. Section 4.2.4.1 o f the TR states that the SSRT tests w ill be conducted at temperatures [ )) at var ious strain rates as described in Tab le 14 o f the TR. The SS RT testing w, econ ucted in Nom ina l Flibe and Redox Cont rolled Flibe to assess if 316H, E R 16 2, and the[~)) of 316H are suscept ib le to env ironmentally ass isted crack ing in Flibe. Sect~ states that the SSRT testing w ill be conducted in acco rdance w ith Amer ican Society fo r Testing and Materials (ASTM ) ASTM G129-0 0, "Standard P ractice for Slo w Stra in Rate Testing to Evaluate the Susceptib ility o f Metallic Mate ria ls to Env ironmentally Assisted Crack ing," 200 0 Ed ition.

In add ition to the SSRT testing, Section 4.2.4.2 of the TR states that fracture mechan ics-based testing of pre-cracked compact tension type samples w ill be used to evaluate fatigue crack g ro wth rates, stress corros ion crack ing rates, and the fractu re behav ior o f 3 16H and ER 16 2 in the Flibe env ironment at temperatures as described in Tab le 15 of the TR.

The test samples w ill cons ist of [

lllln w ith a sha ri:, flaw i.e., the

~ w e ld metal, ea to be tested.

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1 1 1 1 ck w hich w I corros ion testing ) and stress corros ion crack ing portion (a constant stress intens ity factor to initiate stress corrosion under aggress ive testing conditions and then trans ition to cond itions that are representative of the KP-FHR designs ).

The NRC staff finds that the samples to be tested are acceptab le because they are representative of the mater ial and w eldments to be used in the KP-FHR desi ns. In the use of

ros ion om ina l Flibe and Redox Con ro e I e acceptab l....... e in simu lating and determining the crack gro wth rates for

compact tens ion samp les acceptable because these test methods are w ell-established, are an accepted methodo logy, and have been effective in other env ironmentally ass isted cracking testing used by licensees and the NRC staff.

5.1.3.2.2 Env ironmental Creep

Section 4.2.4.3 of the TR describes the mater ial testing that w ill be susce tibili of 316H and ER16-8-2 to env ironmenta l creep in [

)) as described in Table 16 of the TR. The NRC sta m s a pe ormmg creep es mg m oth Norma l Flibe us ing w e lded base metal samp les to inc lude the base meta l, w eld meta l and heat affected zone of the base metal acceptab le since it simulates the mater ia l to be used and the env ironment of the KP-FHR desi ns. In addition, the NRC staff finds it acceptab le that [ )) are not required to be performed un less sign ifican egra a I0n Is no e compare o creep tests performed in a ir, such as fa ilure times outs ide of the 90% confidence interva l from creep tests performed in a ir, or a change in fracture mode because the creep tests in air w ou ld bound the resu lts in Nomina l Flibe in the temperature range of the KP-FHR designs. The NRC staff notes that the creep tests in Nom inal Flibe are to determine if the Flibe contributes add itional degradation beyond those determine from the creep tests performed in air. If the testing determines Flibe has an add itiona l effect on degradation, add itiona l testing w ould be requ ired to quant ify any increase in degradat ion contr ibut ing to Flibe,

and the test resu lts w ould be review ed by the NRC staff in future licensing submittals by the applicant.

5.1.3.3 Metallurgical Effects

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5.1.3.3.1 Stress Relaxat ion Cracking

Section 4.2.5 of the TR states that stress relaxat ion c racking in 316H w ill be addressed by using test results conducted in air up to tempe ratures of [-)) as d iscussed in Section 3.2 of the TR, and by conducting future ana lys is and design r~ents of the KP-FHR designs, such as w e ld designs in Figu re 23 of the TR, and spec ific w eld processes and paramete rs to minim ize stress relaxat ion c racking as detailed in Section 3.3. 1 o f the TR to reduce the triax ia l stresses.

Section 3.3.2 of the TR rov ides

, w hile the heat affect one of 3 16H base meta l w ith

• triaxial str * * *

)).

The NRC staff finds testing in air acceptab le because these test resu lts w ou ld be valid for 316H in Flibe for the KP-FHR designs since triaxial stresses are the ma jor cont ributo r to stress relaxat ion crack ing. In addition, the NRC staff finds that compa ring the suscept ib ility of 316H to that of 347 as d iscussed in Section 3.3.1 o f the T R w ou ld allow a determination o f the bounding triaxia l stresses that cou ld cause stress relaxat ion crac king in 316H. The NR C staff also finds the stress relaxation testing fo r the KP-FHR commerc ia l o w e r reacto r and the non - o w e r test reacto r in Tab le 1 O of the TR acce table because the

3 16H. The NRC staff notes that in Section 3.3.2 * *

• the

• *

• tresses us ing [

)) to better asse idual f 316H for the KP -F HR designs. The resu lts of these ana lys is can be used in futu re licens ing act ions to add ress stress relaxat ion c rac king o f 316H in the KP FHR designs.

5.1.3.3.2 Phase Fo rmation Embrittlement

Section 4.2.5 of the TR d iscusses ho w the qua lification prog ram add resses phase formation embr ittlement, and degradat ion from therma l cycl ing or therma l grad ients. Ka iros states that phase formation embr ittlement may occu r w hen 316H and ER 16-8-2 picks up an e lement dur ing its ex osu re to Flibe and fo rms a de leterious second hase. To address this Ka iros 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 a pplicable to the KP-FHR desi ns that is bo u nd b the test conditions. In this case, a desi n that u tilizes

, d to perform testing to qua n tify the effects on the m echan ical properties of 3 16H and asso ciated w eld fille r meta l ER 16-8-2 (see Limitation and Condi tio n 11 ).

5.1.3.3.3 T hermal Cycl ing / Stripping

T ab le 11 of the TR states that thermal cycl ing [

)]. In add ition, Section 4.2.5 of the TR s a es a egra a I0n o an - -

y g therma l trans ien ts cou ld lead to high stresses resu lting in therma l fatigue degra da tio n.

-Ka iros w ill address the therma l cyclin g b y conduc ting a n alysis to refine the desig n and ope ra tio n of the KP-FHR designs to m itigate large thermal g rad ie n ts. The NRC staff finds it accep ta b le that future ana lys is, in lieu of testin g, w ill be u sed to m itigate thermal cycl ing because the thermal g rad ien ts w ill be m in im ized through the u se of a ppropr ia te design and o pe rating cond itions o f the KP-FHR ( po w er and non - po w e r test reactor ), as info rmed b y the ana lys is.

Ho w ever, since the design has not been fina lized and no testing w ill b e conducted as part of this material qua lification program, the NRC staff is impos ing a limitation and cond ition that an a pplican t imp lement ing this TR w ill add ress therma l cycling / stripping in fu tu re licensing su b m itta ls b y m inimizing the therma l g rad ients v ia ap pro priate des ign and o perating cond itions of KP-FHR designs based on ana lys is (see Lim itation and Condition 12).

5.1.3.4 Irradiation Effects

5.1.3.4.1 Irradiation-Induced Embrittlement

Sec tion 4.2.6.1 o f the TR states tha t existing data ind icates that tens ile properties and fractu re toughness of aus tenitic stainless steels, w hen tested a t high strain rates and tem pera tures from 550°C to 650°C, a re relatively una ffec ted by irrad ia tion levels <0. 1 d isplacemen t pe r atoms (dpa ) w ith a he lium content o f 10 a tomic parts pe r million ( a ppm ) in cu rrent light w a ter reacto r environmen ts. Ho w eve r, a t lo w stra in rates, data sho w s irrad ia tion - induced emb rittlement can affect ma terial properties such as tensile strength and ductility and c reep life due to the gene ra tion o f helium. The a pplican t stated in Section 4.2.6. 1 o f the TR that ex isting da ta w ill be used to dev elo p deg radation facto rs, b u t that it w ill conduct irrad ia tion tests on ER 16 2, 3 16H,

and the associated heat affected zone of 316H to quan tify marg ins at irradiation levels fo r the non-po w er test reac tor and the comme rcial po w e r reacto r w hich w ill be provided in fu tu re licens ing actions. The NRC staff finds it accep ta b le to conduct testing for irrad iation-induced emb rittlement on E R 16-8-2, 316H, and the associated hea t affected zone o f 3 16H, because the testing w ill be representative of the env ironmen t in the KP-FHR des igns and this informat ion w ill b e su b m itted in fu tu re licensing ac tions. N R C staff is impos ing a limitation and cond ition that the test en v ironmen t shall bound the KP-FH R designs, includ ing the expected irradiation damage (dpa ) and he lium con tent (see Limitation and Cond ition 13).

5.1.3.4.2 Irradiation-Affected Corros ion

Section 4.2.6.2 o f the TR sta tes tha t no immed ia te ma teria l testing of 316H and ER 16 2 fo r irradiation effects on cor rosion is pro posed fo r the qua lification of 316H and E R1 6-8-2 b ecause the reac to r vessel has a lo w irrad iation dose level (<0.1 d pa ) and ex isting da ta 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 materials 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 conclusion on the proposed initial planned programs.

Notwithstanding, the NRC staff finds it acceptable 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 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 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 l icense 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 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 2. Consistent with the discussion in Section 5.1.3.4.2 of this SE, above, t his warrants implementation of a materials surveillance program for all commercial

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po w er reactors us ing the KP-FHR design. In addition, the NRC staff finds it acceptab le to use an inspection and mon itoring program to detect cracking of the reactor vessel because the prog ram shou ld be capable of detecting c rack ing that w ou ld prevent the reactor vessel from performing its sa fety funct ion. Therefore, the NRC staff is impos ing a lim itation and condition that the mater ia ls surve illance program and the inspection and monitor ing program w ill be subm itted in future license applications for NRC staff review and approva l to verify that these programs are sufficient to address irradiation - affected stress corros ion crac king of the reactor vessel (see Limitation and Cond ition 14).

5.1.4 Quality Assurance

Section 1 of the TR states that the non-po w er test reactor app lication is implement ing a qua lity assurance p rogram based on ANSI /ANS-15.8 - 1995, "Qua lity Assurance Program Requ irements for Research Reacto rs," (ANS I/ANS-15.8 ), w h ich is endorsed by NRC Regulatory Gu ide 2.5, "Qua lity Assurance Program R equ irements fo r R esearch and Test R eactors. " The NRC staff finds it acceptab le to use ANSI /ANS - 15.8-1995 fo r materia l testing that w ill on ly be used to support the non - po w er test reactor. The N R C staff notes that Rev ision 4 of the T R does not specify if mater ia l testing related to safety-related components w ill be conducted under a program that comp lies w ith the requ irements o f 10 CF R 50 Append ix B, as stated in previous rev isions of the TR. The quality and accuracy of material testing results that cou ld be used for the commercial po w er reactor must be confirmed w hen used to address potent ia l mater ia ls reliability and env ironmenta l compatib ility o f safety-related components. Therefore, the N R C staff is imposing a limitation and cond ition that mate rial testing w ill be conducted under a quality assurance program that complies w ith the requirements of 10 CFR 50 Appendix B to confirm the quality of the data obtained dur ing the mate rial testing that w ill be used for the commercial po w er reactor ( see Limitation and Condition 15).

5.2 Eva luation Summary

The NRC staff finds that the mater ia l qua lification methodo logy for 316H and ER1 6-8-2 materials in Section 4 of the TR satisfy, in part, the PDCs 14 and 3 1 fo r the KP-FHR des igns and is acceptab le, sub ject to the Limitations and Cond itions found in Section 6.0 of the N R C staff's S E be lo w. The NRC staff finds that testing at the norma l ope rating and postu lated accident temperatures, and in both Nomina l Flibe and R edox Controlled Flibe, to va lidate the degradat ion o f 3 16H and E R 16 2 mater ia l, is acceptable since the testing duplicates the operating environment that the mater ia l w ill experience in the KP-FH R designs. The NRC staff a lso finds it acceptable that the material test sam les w ill inc lude not on l the 316H base meta l and associated E R16 2 w e ld meta l but the

e s a aso in s a e re Is reasona e assurance a e egra a I0n mec anisms to be tested as descr ibed in in Section 4 of the T R include the app ropr iate env ironmenta l degradat ion mechanisms for the KP-FHR designs based on the cur rent research and testing informat ion prov ided in the TR and in References 11 and 12 o f this SE. These references discuss topics such as corros ion, env ironmentally ass isted cracking, and the effects o f irradiation on mater ia ls, and their applicab ility in molten sa lt env ironments.

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The staff has reasonable assurance the qualification program meets the requirements listed in Section 4.1 described above, as they relate to the qualification of 316H and ER16 2 in the 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 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, they 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 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.

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6. (Section 4.1 ) As presen ted in the TR, there are key design pa rameters w ithout w hich the proposed reactor coo lant b oundary design and assoc iated properties may no t be su pported. The refo re, K P-FHR designs referencing this TR must have the follow ing :

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

• An intermediate salt coolant loop for the commercia l reactor

• A Primary Hea t Transport System that rejects heat to the a ir in lieu of an intermed iate coolant loop fo r the non-po w er test reac to r

• Non-po w e r test reacto r lifetime of a maximum of 5 years ( 1 year comm ission ing~ opera tion ) and commerc ia l po w er reactor lifetime of a max imum of [-11

• " Near-atmos phe ric" primary coolan t pressu res

• End of life irrad iation of less than 0.1 d pa

These key design parameters of the KP-FHR designs, if changed, cou ld necess itate the mod ification of, or addition to, the testing prog ram.

7.

8. (Section 4.2.3.3 and Table 13) The impur ity effects testing on 316H and ER 16-8-2 must include the po ten tia l loss of Flibe chemistry cont ro l from b oth a ir ingress and intermedia te sa lt loop ingress based on the safety ana lys is reports. An a pplicant referencing this TR mus t demonstra te tha t any poten tia l impur ity ingress (includ ing postulated acc idents ) in the KP-FHR designs is b ound b y the testing performed as part of this TR.

9. (Section 4.2.3.2, Tables 13 and 14) An applicant referenc ing this TR mus t demonstrate tha t the Nom inal Flibe salt com position used in the KP-FHR designs is cons istent w ith the Nom inal Flibe salt composi tion used in these tests in cluding initia l im purities in the sa lt.

10. Section 4.2.3.2, Tables 13 and 14) An a pplicant referencing this top ica l report must demonstrate that the sa lt compositions (w ith reduc ing agen t additions and impuri ties from postu lated acc ident scenar ios ) tested in this prog ram bound any poten tia l salt compos itions for the KP-FHR reacto r designs.

11. (Section 4.2.5) In order to address phase formation embr ittlement for the KP-FHR desi ns an a licant must sho w that testing bounds potentia l design cond itions

[ )) and that if a secondary phase is detected during es mg, e e ec son mec anrca properties of 316H and ER16-8-2 mus t be quan tified v ia testing and app roved by the NRC staff.

12. (Section 4.2.5 and Table 11) The applicant w ill assess thermal cycling / striping in future licensing submitta ls by minim izing the thermal gradients v ia appropr iate design and operating cond itions of the KP-FHR des igns based on analysis.

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13. (Section 4.2.6.1) Testing for irradiation-induced embrittlement of ER16-8-2, 316H, and the associated heat affected zone of 316H must be performed that bounds the environment representative of the KP-FHR designs, including the expected irradiation 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 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 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)

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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) 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 September 2022 (ML21223A097 TBD).

10. US NRC, Regulatory Guide 1.87, Acceptability of ASME Section III, Division 5, High Temperature Reactors, Revision 2, dated September 2022 (ML21091A276 TBD).

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 B eryllium with 316H S tainless Steel in M olten Li2BeF4 (Flibe), Published in Journal of Nuclear Materials, Volume 565, July 2022.

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SUBJ ECT: DRA FT SAFETY EVA LUAT ION O F METALLIC MATE R IA L QUA LIFI CAT ION FO R TH E KA IROS POW ER F LUOR ID E SALT - COO LE D H IGH -T E MP ERATUR E R EACTO R ( K P-TR - 013 ) KAIROS POW ER, LLC EP ID NO. 000431 I 99902069 I L-2020-TO P-0050 DAT ED : D EC EMB ER :XX, 2022

D ISTR IBUT IO N :

NON - PUB LI C RidsACRS_ MailCTR Resou rce R idsNrrDa n uUa l1 R eso ur ce R idsNrrDa n uUa l2 R eso ur ce R idsOgcMa ilCente r Reso u rce R idsOpaMa ilCe n ter R idsOcaMa ilCente r MShams, NR R JBo w e n, N R R CCa rusone, N R R MM itchell, NR R S Philpott, NRR MHayes, NR R WJessup, N RR BBeas ley, NR R SCuad rado, N R R JHonchar ik, NR R AChereskin, N R R RR ivera, N RR DGreene, NR R CS m ith, NRR

ADAMS Accession No. MLXXXXXX.xxxi NRR-106 O FFI CE NRR/DANU / UA L 1/ PM NRR/DA NU / UA L 1/ LA NRR/DN R L/NP HP/ BC NAM E RR ive ra DG reene MM itchell DAT E 12/ / 2022 12/ / 2022 12/ / 2022 O FFI CE OGC NRR/DA NU / UA L 1/ BC NAM E W J essu o DAT E 12/ / 2022 12/ / 2022 OFFICIAL RECORD COPY

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