ML19198A306

Draft Interim Staff Guidance: Supplemental Guidance Regarding the Chromium-Coated Zirconium Alloy Fuel Cladding Accident Tolerant Fuel Concept (ATF-ISG-01)

ML19198A306
Person / Time
Issue date:	07/17/2019
From:	Office of Nuclear Reactor Regulation
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Proffitt J
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Download: ML19198A306 (30)

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1 DRAFT INTERIM STAFF GUIDANCE 2 SUPPLEMENTAL GUIDANCE REGARDING THE 3 CHROMIUM-COATED ZIRCONIUM ALLOY FUEL CLADDING 4 ACCIDENT TOLERANT FUEL CONCEPT 5 ATF-ISG-01 6

7 8 PURPOSE 9

10 The U.S. Nuclear Regulatory Commission (NRC, or Commission) staff is providing this interim 11 staff guidance (ISG) to facilitate the staffs understanding of the in-reactor phenomena important 12 to safety for the chromium-coated zirconium alloy fuel cladding concept being pursued by 13 several U.S. fuel vendors as part of the U.S. Department of Energys accident tolerant fuel 14 (ATF) program.

15 16 BACKGROUND 17 18 This interim staff guidance (ISG) is intended to provide guidance for NRC staff reviewing 19 applications involving fuel products with chromium-coated zirconium alloy cladding. For coated 20 claddings of this type, a phenomena identification and ranking table (PIRT) was generated for 21 the NRC by Pacific Northwest National Laboratory; the guidance provided in this ISG 22 extensively references the PIRT report, Degradation and Failure Phenomena of Accident 23 Tolerant Fuel Concepts: Chromium Coated Zirconium Alloy Cladding (Reference 1). The 24 suggested cladding properties, specified acceptable fuel design limits (SAFDLs), and new 25 failure mechanisms sections from the PIRT are replicated in Appendix B and C, so that 26 modifications to the information may be made based on stakeholder comments and feedback.

27 These appendices supersede sections 5.1 and 5.2 of the PIRT report.

28 This ISG is not intended as stand-alone review guidance, but instead supplements NUREG-29 0800, Standard Review Plan, (SRP, Reference 2) Section 4.2, Fuel System Design, and 30 discusses the potential impact of coated claddings on reviews performed under SRP Section 31 4.3, Nuclear Design, Section 4.4, Thermal and Hydraulic Design, and Chapter 15, Transient 32 and Accident Analysis. In addition to the guidance provided in this ISG, reviewers of coated 33 cladding applications should familiarize themselves with the PIRT report and with the relevant 34 sections of the SRP.

35 The PIRT report and this ISG focus primarily on metallic-chrome coatings applied to a zirconium 36 alloy base metal, with some additional discussion that is applicable to chrome-based ceramic 37 coatings. Reviewers of submittals on ceramic chromium-coated zirconium alloy claddings 38 should carefully read the PIRT to determine the applicability to the review.

ATF-ISG-01 Page 2 of 4 39 40 This ISG does not apply to reviews of fuel products other than metallic or ceramic chromium-41 based coatings on a zirconium alloy substrate.

42 RATIONALE 43 44 The current review guidance in the SRP assumes the use of uranium dioxide fuel pellets 45 contained within zirconium alloy-based fuel cladding and is targeted to specific degradation and 46 failure modes associated with that material. Based on this fact, along with the aggressive 47 development timelines of DOE and industry ATF programs, the staff proactively developed a 48 plan, Project Plan to Prepare the U.S. Nuclear Regulatory Commission for Efficient and 49 Effective Licensing of Accident Tolerant Fuels (ATF Project Plan, Reference 3) to outline a 50 preparation strategy for ensuring staff readiness to perform timely licensing reviews. This ISG 51 will serve as the concept-specific licensing roadmap for chromium-coated zirconium alloy 52 cladding that is detailed as part of the strategy included in the ATF Project Plan.

53 54 APPLICABILITY 55 56 This guidance applies to:

57 58 All holders of an operating license or construction permit for a nuclear power reactor under Title 59 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic Licensing of Production and 60 Utilization Facilities, except those who have permanently ceased operations and have certified 61 that fuel has been permanently removed from the reactor vessel.

62 63 All holders of and applicants for a power reactor early site permit, combined license, standard 64 design approval, or manufacturing license under 10 CFR Part 52, Licenses, Certifications, and 65 Approvals for Nuclear Power Plants. All applicants for a standard design certification, including 66 such applicants after initial issuance of a design certification rule.

67 68 All holders of and applicants for a power reactor early site permit (ESP), combined license 69 (COL), standard design certification (DC), standard design approval (DA), or manufacturing 70 license (ML) referencing a small modular reactor (SMR) design under Title 10 of the Code of 71 Federal Regulations (10 CFR) Part 52, Licenses, Certifications, and Approvals for Nuclear 72 Power Plants. SMRs are defined using the International Atomic Energy Agency definition of 73 small and medium-sized reactors with an electrical output of less than 700 megawatts.

74 75 All contractors and vendors (C/Vs) that supply basic components to U.S. Nuclear Regulatory 76 Commission (NRC) licensees under Title 10 of the Code of Federal Regulations (10 CFR) Part 77 50, Domestic Licensing of Production and Utilization Facilities or 10 CFR Part 52, Licenses, 78 Certifications, and Approvals for Nuclear Power Plants.

79

ATF-ISG-01 Page 3 of 4 80 GUIDANCE 81 82 The information contained in Appendix A to this ISG provides supplemental guidance to 83 Chapters 4 and 15 of the SRP for NRC reviewers. The foundation for this additional guidance is 84 the chromium-coated cladding PIRT report. Reviewers should ensure that applicants adequately 85 address or disposition each of the criteria cited in the guidance as appropriate for the specific 86 chromium coated cladding technology in reaching a reasonable assurance conclusion.

87 88 IMPLEMENTATION 89 90 The staff will use the information contained in this ISG to ensure that all known degradation and 91 failure mechanisms for chromium-coated zirconium alloy fuel cladding are considered such that 92 their impact on the acceptance criteria contained in SRP sections 4.2, 4.3, and 4.4 along with 93 chapter 15 can be assessed.

94 95 BACKFITTING AND ISSUE FINALITY DISCUSSION 96 97 Discussion to be provided in final ISG.

98 99 CONGRESSIONAL REVIEW ACT 100 101 Discussion to be provided in final ISG.

102 103 FINAL RESOLUTION 104 105 By 2025, this information will be transitioned into Chapters 4 and 15 of the SRP. Following the 106 transition of this guidance to the SRP, this ISG will be closed.

107 108 109 APPENDICES 110 111 A. Supplemental Guidance for SRP Chapters 4 and 15 112 B. Cladding Material Property Correlations 113 C. Specified Acceptable Fuel Design Limits (SAFDLs) 114 D. (Placeholder) Resolution of Public Comments 115 116 REFERENCES 117 118 1. Chromium-Coated Cladding Final PIRT Report, Degradation and Failure Phenomena of 119 Accident Tolerant Fuel Concepts: Chromium Coated Zirconium Alloy Cladding, June 120 2019 (Agencywide Documents Access and Management System (ADAMS) Accession 121 No. ML19172A154)

ATF-ISG-01 Page 4 of 4 122 2. NRCs Standard Review Plan, Standard Review Plan for the Review of Safety Analysis 123 Reports for Nuclear Power Plants: LWR Edition (NUREG-0800), (ADAMS Accession 124 No. ML070810350) 125 3. NRCs ATF Project Plan, Project Plan to Prepare the U.S. Nuclear Regulatory 126 Commission for Efficient and Effective Licensing of Accident Tolerant Fuels, September 127 2019 (ADAMS Accession No. ML18261A414) 128 129 Public Meetings: August 6, 2019; December 4, 2019 130

1 APPENDIX A 2

3 Supplemental Guidance for SRP Chapters 4 and 15 4

5 NUREG-0800 - Chapter 4, Section 4.2, Fuel System Design 6 For reviews of new fuel products where the only change from an existing approved fuel design 7 that utilizes zirconium alloy cladding is the adoption of chromium-coated cladding, the licensing 8 of a new cladding alloy can be used as a model. While SRP 4.2 covers additional requirements 9 for review of complete fuel systems, cladding reviews cover these three areas:

10

• Definition of specified acceptable fuel design limits (SAFDLs) for new cladding, 11

• Material property correlations to be used in codes to ensure the new cladding satisfies 12 the SAFDLs, and 13

• Any changes that must be made to existing methodologies to accommodate the new 14 cladding.

15 These topics will be discussed in more detail in the following sections.

16 While chromium coatings may only be a fraction of the thickness of the base cladding, they are 17 designed to provide the following benefits over uncoated cladding:

18

• Harder surface 19 o Improves cladding fretting performance and wear resistance 20 21

• Negligible oxidation during normal operation 22 o Protects zirconium cladding from oxidation 23 o Protects zirconium cladding from hydrogen uptake 24 25

• Improved high temperature steam oxidation kinetics 26 o Reduced rate of corrosion and heat of oxidation 27 o Protects zirconium cladding from oxidation 28 o Reduced hydrogen liberation 29 30

• Improved high temperature strength 31 This ISG does not attempt to set standards for review of any credit or benefit applicants may 32 request by demonstrating these improvements, as strategies for licensing these potential 33 improvements have not yet been submitted to the Nuclear Regulatory Commission. The 34 reviewer of any coated cladding must, therefore, evaluate any proposed property improvements 35 against the data provided by the applicant. The reviewer must also evaluate if the data provided 36 supports the full operating domain for the fuel, and place appropriate limitations and conditions 37 when necessary. Finally, if an applicant wishes to take credit for coating behavior up to a certain 38 burnup, or during certain accident conditions, it is necessary for the adherence of that coating to 39 the substrate to have been demonstrated to that burnup and during those conditions.

ATF-ISG-01, Appendix A Page 2 of 7 40 Definition of SAFDLs for New Cladding 41 The SAFDLs mentioned in SRP Section 4.2 under SRP Acceptance Criteria, Design Bases 42 can be broadly separated into three general categories:

43

• SAFDLs related to fuel assembly performance that are typically addressed by simple 44 calculation, manufacturing controls, and historical data 45

• SAFDLs related to fuel rod performance that are typically addressed for normal 46 operation and anticipated operational occurrences (AOOs) using a thermal mechanical 47 code 48

• SAFDLs related to fuel rod performance that are typically addressed for accident 49 conditions using a system analysis code with initial conditions provided by a thermal 50 mechanical code.

51 Each SAFDL listed in SRP 4.2 is included in Table 5.2 of the PIRT report and described in 52 further detail in Appendix C of this ISG. These sections detail the expected and potential impact 53 of the coatings on each SAFDL.

54 The reviewer should ensure that chromium-coated cladding submittals address each of the 55 SAFDLs where the PIRT report notes that additional concerns may exist. Table 5.3 of the PIRT 56 report contains a summary of tests that could be performed to justify SAFDLs; however, the 57 NRC does not require any specific testing to be performed and applicants may be able to 58 sufficiently address a SAFDL in an alternate fashion. If a submittal is under review, some of the 59 SAFDLs may be left to address in application-specific reviews, as plants apply for license 60 amendments to load batch quantities of fuel with coated cladding. If this is the case, these 61 should be noted in the safety evaluation for the application for the coated cladding product, 62 typically as a condition or limitation.

63 Potential new damage mechanisms have been identified in Appendix C, Section C.4 of this ISG.

64 The reviewer should ensure that these mechanisms have been ruled out sufficiently by the 65 applicant for the domain approved by the NRC, that existing SAFDLs already protect against the 66 mechanisms, or that new SAFDLs have been developed to protect against them.

67 Based upon an investigation of available performance testing and known data gaps, Section 68 6.4.2 of the PIRT report identified several performance concerns for chromium-coated zirconium 69 alloys. The reviewer should ensure that these performance concerns have been ruled out 70 sufficiently by the applicant for the domain approved by the NRC, that existing SAFDLs already 71 protect against the damage mechanisms, or that new SAFDLs have been developed to protect 72 against them.

73 With respect to LOCA post-quench ductility (PQD), the PIRT report identifies that the 10 CFR 74 50.46 regulatory limits of 2200°F (1204°C) peak cladding temperature (PCT), and 17%

75 equivalent cladding reacted (ECR) maximum local oxidation are likely inappropriate 76 embrittlement limits for chromium-coated zirconium alloys. These analytical limits for PQD were 77 based on ring compression tests (RCT) conducted on zirconium cladding segments exposed to 78 various levels of high temperature steam oxidation. The point of nil-ductility was predicted by

ATF-ISG-01, Appendix A Page 3 of 7 79 integrating time-at-temperature using the Baker-Just weight gain correlation. Embrittlement of 80 the cladding is governed by oxygen diffusion into the base metal. Though highly correlated for 81 uncoated zirconium alloys, the amount of cladding outer surface oxidation (i.e., measured ECR) 82 is not the direct cause of cladding embrittlement. Differences in the oxidation kinetics between 83 zirconium-based cladding and chromium-coated cladding will challenge both the existing 17%

84 ECR analytical limit based on Baker-Just and, more generally, the use of maximum local 85 oxidation (i.e., predicted ECR) as a surrogate SAFDL for cladding embrittlement due to oxygen 86 diffusion. This issue is highlighted in Section 6.2.6 of the PIRT report, which describes 87 chromium-coated zirconium alloy cladding loss-of-coolant-accident (LOCA) PQD test results 88 showing significant differences between allowable predicted ECR (beyond 17%) and measured 89 ECR at nil-ductlity (3-5%). Note that Section 6.2.6 of the PIRT report incorrectly refers to hydride 90 embrittlement instead of oxygen diffusion-based embrittlement.

91 Section 4 of the PIRT report describes the zirconium-chromium phase diagram. The formation 92 of a liquid phase at the eutectic point shown at 1332°C, which is well below the melting point of 93 either the chromium coating or the zirconium alloy substrate, is another concern with respect to 94 establishing a PCT SAFDL. The reviewer should ensure that the applicant provides a sufficient 95 empirical database to define performance metrics and analytical limits which preserve 96 acceptable fuel rod behavior under LOCA conditions.

97 As described in Section 6.2.2 of the PIRT report, chromium coating may also impact the fuel rod 98 ballooning characteristics under accident conditions. While no regulatory limits are currently 99 defined to limit the extent of ballooning or the size of the rupture opening, concerns related to 100 fuel fragmentation, relocation, and dispersal may warrant future SAFDLs for fuel rod burnup 101 extensions beyond rod-average values of 62 GWd/MTU.

102 Material Property Correlations to Ensure SAFDLs are Met 103 Appendix B provides a list of cladding material properties that are typically needed to 104 adequately model fuel system response based on development and qualification of NRCs 105 independent fuel performance code, FRAPCON, and previously approved thermal-mechanical 106 codes. These property correlations are then used by the thermal-mechanical codes to 107 demonstrate compliance with the SAFDLs. This approval may come at the topical report review 108 stage, if an applicant demonstrates that the SAFDL is satisfied for the entire design and 109 operating domain, or a methodology may be approved to be used for each licensee that wishes 110 to load the fuel.

111 The PIRT report also suggests two paths that an applicant may take to analyze each property:

112 treating the cladding and coating as separate layers and treating the cladding and coating 113 together as a composite material. A subset of the composite material strategy may be to ignore 114 the coating (for the purposes of thermal-mechanical analyses) and use the properties of the 115 underlying cladding substrate. Any of these paths may be appropriate provided sufficient 116 justification from the applicant, and a variety of these strategies may be used to disposition the 117 various properties.

ATF-ISG-01, Appendix A Page 4 of 7 118 Appendix B details each of the twelve properties identified in the report. Applicants intending to 119 use chromium-coated zirconium alloy cladding should address all these properties. If the 120 applicant assumes that the coated cladding will behave the same as the underlying substrate 121 without supporting evidence that the property is unchanged, this assumption should be 122 demonstrated to be conservative for normal operation, AOOs, and accidents described in 123 Section 15 of the SRP.

124 Changes to Existing Codes and Methodologies 125 New cladding properties need to be properly modeled using computer codes to assess the 126 performance of the coated cladding. If, for a given property, the coated cladding is treated as a 127 composite material, changes to the codes and methods may not be needed beyond updates to 128 the property correlations; however, if the cladding is treated as a separate layer, codes may 129 need to be modified to account for the additional layer as well as interface effects.

130 Regardless of the changes made to address the coating, the codes and methods must be 131 validated. Section 5.3.1 of the PIRT report identifies five areas where validation is critical:

132

• Fuel temperature 133

• Fission gas release 134

• Rod internal pressure and void volume 135

• Cladding oxide thickness 136

• Cladding permanent hoop strain following a power ramp.

137 Sections 5.3.1.1 through 5.3.1.5 of the PIRT report go into each of these in more detail. Table 138 5.4 of the PIRT report provides a list of test data that may be used in code assessment.

139 The methodology for performing the fuel system safety analysis consists of the following pieces:

140

• Identification of functional requirements for the fuel and assembly 141

• Identification of limits for each functional requirement 142

• Identification of code or other approach that will be used to assess performance against 143 functional requirement 144

• Identification of approach to demonstrate high level of confidence that design will not 145 exceed functional requirements:

146 o Selection of power histories to be considered 147 o Identification of uncertainties in operational parameters 148 o Identification of fabrication uncertainties 149 o Identification of modeling uncertainties 150 o Approach to quantify an upper tolerance level based on identified uncertainties.

151 The identification of functional requirements for the fuel and assembly and the limits for each is 152 satisfied by the selection of appropriate SAFDLs. There have been new damage mechanisms 153 identified in Appendix C, Section C.4, that should be implicitly handled via existing SAFDLs and 154 considered in the development of those SAFDL limits. Alternatively, the methodology may be

ATF-ISG-01, Appendix A Page 5 of 7 155 modified to explicitly address these mechanisms through new functional requirements and 156 limits.

157 The material property updates and the code assessment have been discussed. No further 158 methodology change is anticipated as far as the use of codes is considered. The identification of 159 operational parameters such as rod power, coolant flow rate, etc. is not expected to be 160 impacted by the implementation of chromium-coated zirconium alloy cladding. Any further 161 changes to the code or operational parameters should be evaluated during the review of the 162 application.

163 The identification of fabrication uncertainties will be taken from uncertainty specifications on the 164 drawings or from manufacturing data. Although specific values may change, the general 165 approach for obtaining these values is not expected to change. Any changes to this general 166 approach should be dispositioned sufficiently in the application.

167 Modeling uncertainties should be identified during the implementation and assessment of new 168 material properties in codes. Comparing property data to correlations and code predictions to 169 measurements should allow for the appropriate development of acceptable modeling 170 uncertainties. The application should identify modeling uncertainties and explain how the 171 uncertainties were determined.

172 Existing approaches to calculate upper tolerance levels are robust and should be acceptable to 173 perform these calculations for chromium-coated zirconium alloy cladding assuming that the 174 activities discussed above are rigorously performed. Any changes to these approaches should 175 be dispositioned in the application.

176 NUREG-0800 - Chapter 4, Section 4.3, Nuclear Design 177 Section 4.3 of NUREG-0800, (the Standard Review Plan) covers the review of the nuclear 178 design of fuel assemblies, control systems, and the reactor core. The reviewer of coated 179 cladding in this area should ensure that the cross-sections generated for the fuel include the 180 effect of the coating.

181 NUREG-0800 - Chapter 4, Section 4.4, Thermal and Hydraulic Design 182 Section 4.4 of NUREG-0800 covers the thermal hydraulic design for fuel assemblies, including 183 critical heat flux (CHF) or critical power ratio correlations. The reviewer of a coated cladding 184 submittal in this area should ensure that the following areas are addressed:

185 1) Changes to hydraulic diameter due to the coating thickness 186 2) Changes to boiling crisis behavior, including effects of surface roughness 187 3) For boiling water reactor applications, changes to rewet temperature following dryout 188 (i.e. Tmin) 189 Coating degradation mechanisms, as discussed in Appendix C of this ISG, may affect the 190 cladding thermal-hydraulic characteristics. This is particularly true for coating cracking and 191 delamination, which have the potential to change the flow and/or boiling regime near the

ATF-ISG-01, Appendix A Page 6 of 7 192 cladding surface. Coating cracking and delamination may also result in nucleation sites that 193 have the potential to cause hot spots and localized corrosion. The reviewer should ensure that 194 these effects are appropriately accounted for or that coating degradation is otherwise prevented.

195 NUREG-0800 - Chapter 15, Transient and Accident Analyses 196 USFAR Chapter 15 provides demonstration that the Technical Specification (TS) Limiting 197 Conditions of Operation, TS Limiting Safety System Setting, and Reactor Protection System and 198 Engineered Safety Features Actuation System are capable of performing their safety functions, 199 ensuring fuel does not exceed SAFDLs during normal operation and AOOs, and mitigating the 200 consequences of postulated accidents. Chapter 15 of NUREG-0800 provides guidance for the 201 review of these safety analyses.

202 As described above for SRP Section 4.2, chromium coatings may impact the claddings material 203 properties and mechanical and thermal behavior. These changes should be incorporated, where 204 necessary, in the fuel rod thermal-mechanical models which provide important fuel parameters 205 and initial conditions to the reactor core neutronic (SRP Section 4.3) and thermal-hydraulic 206 (SRP Section 4.4) models and nuclear steam supply system codes used in the Chapter 15 207 demonstration.

208 Chromium coatings may have an impact on the cladding initial condition and mechanical 209 properties at the onset of AOOs and postulated accidents. Depending on the oxidation 210 characteristics of the chromium coated cladding, the load-bearing zirconium cladding may 211 experience little-to-no corrosion-related wall thinning and potentially less hydrogen uptake. This 212 reduces cladding stress and preserves beneficial ductility prior to a transient event. AOO 213 overpower cladding strain analytical limits, reactivity-initiated accident pellet-cladding 214 mechanical interaction (RIA PCMI) cladding failure thresholds (See DG-1327), and LOCA PCT 215 and integral time-at-temperature analytical limits (See rulemaking on 10 CFR 50.46c) are all 216 influenced by initial cladding hydrogen content. Hence, any reduction in hydrogen uptake 217 provided by the chromium coating would have a beneficial impact for these transient events.

218 As described above for SRP Section 4.2, the addition of a chromium coating may necessitate 219 changes to existing SAFDLs or require new SAFDLs. These impacts would need to be 220 incorporated into the Chapter 15 demonstration.

221 Any inherent impacts of the chromium coating which potentially impact the fuel rod initial 222 conditions (e.g., gap conductivity, stored energy) should be captured in the fuel rod performance 223 models (SRP Section 4.2). Similarly, potential impacts on core reactivity should be captured in 224 the reactor physics models (SRP Section 4.3). Finally, potential impacts on the rod-to-coolant 225 heat transfer, CHF correlation and safety limits should be captured in core TH models (SRP 226 Section 4.4).

227 For many AOOs and postulated accidents, the presence of a thin chromium coating is not 228 expected to play a significant role on the fuel rods performance during the transient nor 229 influence the overall accident progression. For example, PWR UFSAR Chapter 15.2 safety 230 analyses demonstrates that over-pressure protection systems (e.g., main steam safety valves, 231 pressurizer safety valves) protect the integrity of the reactor pressure boundary during decrease

ATF-ISG-01, Appendix A Page 7 of 7 232 in secondary heat removal AOOs and postulated accidents. For this demonstration, the fuel 233 rods are not modelled in specific detail and the presence of a thin chromium coating will have no 234 impact.

235 For AOOs and postulated accidents involving an increase in global or local core power (for 236 example PWR excess steam demand or main steam line break, BWR loss of feedwater heater 237 or turbine trip, PWR inadvertent bank withdrawal or control rod ejection, and BWR rod 238 withdrawal error or blade drop), the presence of a brittle chromium coating may act as a 239 nucleation site for crack propagation into the base zirconium cladding. Alternatively, a thin 240 ductile chromium coating would likely not initiate crack propagation. A review of coated cladding 241 products under SRP Section 4.2 should evaluate the potential impact of the chromium coating 242 on the claddings strain loading capability and whether a revised AOO overpower cladding strain 243 failure threshold (e.g. 1.0% permanent) or revised RIA PCMI cladding failure thresholds is 244 needed. Nevertheless, the presence of the chromium coating will not change the systems 245 response to the initiating event.

246 For AOOs and postulated accidents involving a decrease in reactor coolant flow (for example, 247 loss of A/C power and PWR reactor coolant pump locked rotor), the presence of the chromium 248 coating will not change the systems response to the initiating event.

249 During a postulated LOCA, the design features of the chromium coating are expected to have 250 an impact on the fuel rods performance during the transient. During the LOCA, multiple 251 parameters may be affected as presented in the table below. The predicted PCT will likely be 252 reduced (due to reduced heat of oxidation), oxygen ingress to the cladding outside diameter will 253 likely be reduced (due to reduction in initial source, rate of oxidation, and lower PCT), hydrogen-254 enhanced beta-layer embrittlement will likely be reduced (due to lower initial cladding hydrogen 255 content), and plastic strains may be reduced (due to coating high temperature strength).

256 As a result of these improvements, chromium-coated fuel rod structural integrity and coolable 257 geometry may be more readily maintained than with a typical, uncoated zirconium-alloy-based 258 cladding.

259 While it is not expected that the chromium coating will improve fuel rod cladding-to-coolant heat 260 transfer, LOCA core temperatures may be reduced due to the reduction in heat addition from 261 cladding oxidation. These lower temperatures, combined with improved oxidation kinetics, will 262 reduce core wide inventories of liberated hydrogen.

263 The reviewer should ensure that the impact of chromium coating on each of the Chapter 15 264 AOOs and postulated accidents has been properly assessed. The scope of work needed to 265 complete the Chapter 15 demonstration increases significantly if the chromium coating 266 negatively impacts fuel temperature, fuel rod cladding-to-coolant heat transfer or CHF 267 correlation or if the application is accompanied with an increase in fuel rod peaking factors, 268 cycle length, allowable fuel rod burnup, or increased 235U enrichment.

269

1 APPENDIX B 2

3 Cladding Material Property Correlations 4

5 The following cladding material properties are typically needed to perform fuel thermal-6 mechanical analysis of nuclear fuel with Zr-alloy cladding under normal conditions and AOOs:

7

• thermal conductivity 8

• thermal expansion 9

• emissivity 10

• enthalpy and specific heat 11

• elastic modulus 12

• yield stress 13

• thermal and irradiation creep rate (function of stress, temperature, and fast neutron flux) 14

• axial irradiation growth 15

• oxidation rate 16

• hydrogen pickup.

17 The following additional material properties are typically needed to perform fuel-mechanical 18 analysis of nuclear fuel under accident conditions based on the development and qualification of 19 the NRC transient fuel performance code, FRAPTRAN (Geelhood K. , Luscher, Cuta, & Porter, 20 2016):

21

• High temperature ballooning behavior 22

• High temperature (800-1200°C) steam oxidation rate.

23 If the first approach discussed above to independently model the coating and the cladding is 24 taken, then each of the above properties and the impact of irradiation on these should be 25 determined as well as the interface behavior. If the second approach discussed above to model 26 the cladding and the coating as a composite material is taken, then the impact of the coating on 27 the base metal should be determined. The following discussion provides information on the 28 potential impact of a metallic or ceramic coating on the base metal.

29 Each of these properties are discussed in the following sections as they relate to Cr-coated Zr 30 cladding. The type of data that are typically used to justify each property will be stated. Currently 31 it is not possible to definitively state what data are available to justify these properties, because 32 small differences in vendor specific processes can have a significant impact on the properties.

33 Therefore, the applicant should provide data or other justification from their specific cladding 34 product to justify material property models. There is a growing body of generic data from various 35 Cr-coated Zr samples as discussed in Section 6.0 of the PIRT report. These data are important 36 because they provide the NRC staff a baseline of what to expect when reviewing an application 37 and claims of large deviations from the generic database may indicate an area for a more 38 detailed review. In the following discussion it should be noted that the coatings under 39 consideration are 5 to 30 microns thick on cladding that is 500 to 700 microns thick. Table 5.1 in

ATF-ISG-01, Appendix B Page 2 of 7 40 the PIRT report provides a summary of the tests that could be performed to quantify the material 41 properties discussed below.

42 B.1: Thermal Conductivity 43 Zr-alloy Cladding 44 Cladding thermal conductivity is not expected to change significantly with irradiation based on 45 the currently available data. Typically heat transfer in a metal is due to electronic heat transfer 46 which is not significantly impacted by lattice damage done by fast neutron irradiation. No 47 change in thermal conductivity with irradiation is used in FRAPCON (Luscher, Geelhood, &

48 Porter, 2015). Thermal conductivity data as a function of temperature from unirradiated samples 49 have typically been used to develop cladding thermal conductivity correlations.

50 Cr-coated Zr 51 Either an effective thermal conductivity for the coated cladding could be developed or a method 52 for combining the thermal conductivity from the base metal and the coating could be described.

53 The thermal conductivity of Cr metal is not expected to be strongly impacted by irradiation. The 54 thermal conductivity of a Cr-based ceramic may be impacted by irradiation. It is possible that the 55 overall cladding thermal conductivity may not be strongly impacted by this as the coating is 56 expected to be relatively thin. However, a ceramic coating will have a greater impact as the 57 thermal conductivity of ceramics are generally low. This would be similar to the treatment of the 58 ZrO2 that evolves on the surface of the Zr-alloy cladding.

59 B.2: Thermal Expansion 60 Zr-alloy Cladding 61 Cladding thermal expansion is not expected to change significantly with irradiation based on the 62 currently available data. Thermal expansion is caused by crystal lattice expansion and does not 63 change much with the introduction of dislocations from fast neutron irradiation. No change in 64 thermal expansion with irradiation is used in FRAPCON (Luscher, Geelhood, & Porter, 2015).

65 Thermal expansion data as a function of temperature from unirradiated samples have typically 66 been used to develop cladding thermal expansion correlations.

67 Cr-coated Zr 68 Typically, the thermal expansion of a coated part will be the same as that of an uncoated part if 69 the coating is relatively thin. However, thermal expansion data from representative cladding 70 tubes would be useful to justify the correlation and to demonstrate that there has not been a 71 change in behavior with the coating due to thermal expansion mismatch between the substrate 72 and the coating. Thermal expansion mismatch between a coating and substrate typically results 73 in plastic strain in the thin coating which is weaker than the substrate because of its thickness.

74 This is particularly true for the Zr-Cr system since the textured hexagonal crystal structure leads 75 to different thermal expansion in different directions, while the cubic Cr or Cr-ceramic coatings 76 will have similar thermal expansion in all directions. Many ceramics have a limited strain 77 capability. A ceramic coating with a significant thermal expansion mismatch strain may exhibit 78 cracking upon heating and cooling due to the inability of that coating to tolerate plastic strain.

79 Application methods may also lead to different thermal expansion mismatch. For example, 80 electroplated coatings can usually not tolerate large strains, PVD coatings are usually dense 81 and adherent, and plasma spray coatings can result in anisotropic mechanical properties due to

ATF-ISG-01, Appendix B Page 3 of 7 82 the spray direction, i.e., in plane versus out of plane property differences. The effects of thermal 83 expansion mis-match and their inherent interface strains can be mitigated by processing 84 conditions. For instance, surface treatments that enhance surface area, strain tolerant 85 microstructures, and higher ductility compliant layers can be utilized to reduce interface strains.

86 B.3: Emissivity 87 Zr-alloy Cladding 88 Cladding emissivity is important to calculate the portion of the gap heat transfer due to radiative 89 heat transfer. The emissivity is impacted by the surface conditions including any oxide on the 90 surface of the cladding.

91 Cr-coated Zr 92 The gap heat transfer occurs on the inner surface of the tube and will not be impacted by the 93 coating on the outer surface. Some system codes and accident analysis codes account for 94 cladding surface emissivity and radiation heat transfer from fuel rods to other reactor core 95 components. The outer surface emissivity may be important in severe accident analysis or even 96 in design basis accident analysis (especially if licensees propose higher peak cladding 97 temperature limits for their plants). Because the current coatings are on the outer surface it 98 would be acceptable to retain the emissivity used for an uncoated Zr-alloy tube for thermal-99 mechanical analysis, but it may be necessary to revise the outer surface emissivity for accident 100 analyses. This would apply equally to metallic and ceramic coatings. (Seshadri, Philips, &

101 Shirvan, 2018) 102 B.4: Enthalpy and Specific heat 103 Zr-alloy Cladding 104 Cladding enthalpy and specific heat are not expected to change significantly with irradiation 105 based on the currently available data. Specific heat of a material is dependent on the 106 composition and the crystal structure and does not change much with the introduction of 107 dislocations from fast neutron irradiation. No change in enthalpy or specific heat with irradiation 108 is used in FRAPCON (Luscher, Geelhood, & Porter, 2015). Enthalpy and/or specific heat data 109 as a function of temperature from unirradiated samples would be useful to develop cladding 110 enthalpy and specific heat correlations.

111 Cr-coated Zr 112 Either an effective enthalpy and specific heat for the coated cladding could be developed or a 113 method for combining the enthalpy and specific heat from the base metal and the coating could 114 be described. Cladding enthalpy and specific heat are only needed for transient fuel 115 performance analysis and for calculation of stored energy. This would apply equally to metallic 116 and ceramic coatings.

117 B.5: Elastic Modulus 118 Zr-alloy Cladding 119 Cladding elastic modulus has been observed to be a weak function of fast neutron fluence 120 (proportional to fuel burnup) (Geelhood, Beyer, & Luscher, PNNL Stress/Strain Correlation for 121 Zircaloy. PNNL-17700, 2008). Not all applicants include a fluence dependence, but if one is

ATF-ISG-01, Appendix B Page 4 of 7 122 included, then temperature dependent data from irradiated and unirradiated coated tubes would 123 be useful to justify the correlation used.

124 Cr-coated Zr 125 Recent data on unirradiated Cr-coated Zr indicate the elastic modulus of a coated part will be 126 the same as that of an uncoated part (Brachet, et al., 2017) (Kim, et al., 2015) (Shahin, Petrik, 127 Seshadri, Phillips, & Shirvan, 2018). Typically, ceramic materials are stiffer (greater elastic 128 modulus) than metallic materials. However, for thin coatings the enhanced stiffness of the 129 coating is not expected to strongly impact the overall stiffness of the substrate. Nano-indentation 130 could be used to evaluate the elastic modulus of the coating.

131 B.6: Yield Stress 132 Zr-alloy Cladding 133 Cladding yield stress has been observed to be a strong function of fast neutron fluence 134 (proportional to fuel burnup) early in life and saturates to a value at moderate fluence levels.

135 Temperature dependent data from irradiated and unirradiated coated tubes should be provided 136 to justify the correlation used.

137 Cr-coated Zr 138 Recent data on unirradiated Cr-coated Zr indicate the yield stress of a coated part will be the 139 same as that of an uncoated part (Brachet, et al., 2017) (Kim, et al., 2015) (Shahin, Petrik, 140 Seshadri, Phillips, & Shirvan, 2018). In tension, ceramic materials display a wide variation in 141 strength. However, for thin coatings the variable strength of the coating is not expected to 142 strongly impact the overall strength of the substrate. Nano-indentation could be used to evaluate 143 the yield stress of the coating. Although the yield stress of the tube may not change, if the 144 thickness of the substrate tube is reduced to accommodate a coating that offers no strength, 145 then the maximum load capability of that tube will be reduced. Generally, coating is assumed 146 not to offer any load bearing capability.

147 B.7: Thermal and Irradiation Creep Rate 148 Zr-alloy Cladding 149 The creep behavior of zirconium alloy tubes has often been characterized by a thermal rate 150 which can be developed based on ex-reactor creep tests, which are a function of stress and 151 temperature, and an irradiation rate which can be developed based on the additional creep 152 observed at the same stress and temperature during an in-reactor creep test. This creep rate 153 can change significantly with small changes to alloy composition or microstructure. The increase 154 or decrease in the thermal creep rate does not directly correlate to an increase or decrease in 155 the irradiation creep rate. One example of this is the creep rates for recrystallized cladding and 156 stress-relief annealed cladding in FRAPCON. Although both the thermal and irradiation creep 157 rates are greater for the stress-relief annealed cladding than the recrystallized cladding, the two 158 increases are not the same fraction so one increase could not be determined from the other 159 (Geelhood K. , Luscher, Raynaud, & I.E., 2015) (Limback & Andersson, 1996). Both in-reactor 160 and ex-reactor creep tests are recommended to justify the cladding creep correlation used as 161 these processes are potentially controlled by different mechanisms.

162

ATF-ISG-01, Appendix B Page 5 of 7 163 Cr-coated Zr 164 Recent data on unirradiated Cr-coated Zr indicate the thermal creep behavior of a coated part 165 will be the same as that of an uncoated part (Brachet, et al., 2017). A thin metallic or ceramic 166 coating on the cladding is unlikely to impact the thermal or irradiation creep behavior of the 167 substrate. However, as mentioned above, small changes in composition and microstructure can 168 have a significant impact on creep behavior, such that the application of the metallic or ceramic 169 coating may impact the creep behavior. For this reason, both in-reactor and ex-reactor creep 170 tests are recommended to justify the cladding creep correlation used for Cr-coated Zr cladding.

171 The coating will put the substrate under compression (depending on methodology) which may 172 improve the creep properties.

173 B.8: Axial Irradiation Growth 174 Zr-alloy Cladding 175 Zirconium alloy tubes have been observed to grow axially with increased fast neutron fluence 176 (Luscher, Geelhood, & Porter, 2015). This growth rate can change significantly with small 177 changes to alloy composition, texture, or microstructure (for example, Zircaloy-2, Zircaloy-4, 178 M5, ZIRLO). In-reactor data would be useful to justify the axial growth correlation used.

179 Cr-coated Zr 180 There is no current experience with the axial irradiation growth of coated parts relative to 181 uncoated parts. Like thermal expansion mismatch strain, a difference in growth rates between 182 the coating and substrate could lead to plastic deformation in the coating. This could be 183 especially exacerbated for ceramic coatings as ceramics typically have low plastic strain 184 capability. Large differences in growth rate between the cladding and coating could lead to 185 cracking or adhesion issues.

186 B.9: Oxidation Rate 187 Zr-alloy Cladding 188 The oxidation rate is important to model in uncoated cladding tubes as the zirconium oxide layer 189 is less conductive than Zr metal. In the zirconium alloy systems, ex-reactor autoclave corrosion 190 data is significantly different from in-reactor corrosion data and should not be used to develop 191 corrosion correlations for coated parts. Additionally, the corrosion behavior of non-fueled 192 cladding segments may also not be representative of fueled cladding corrosion as the surface 193 heat flux in the fueled cladding seems to strongly impact oxidation rate (Cox, 2005) (Sabol, 194 Comstock, Weiner, Larouere, & Stanutz, 1993) (Garde, Pati, Krammen, Smith, & Endter, 1993).

195 Cr-coated Zr 196 The Cr coatings under consideration will most likely result in very low oxidation rates under 197 normal conditions and AOOs. Both the metallic and ceramic Cr coatings tend to produce a 198 protective chromium oxide layer that exhibits excellent corrosion resistance, but this is a 199 function of the coating application method. Some in-reactor data from fueled rods under 200 prototypical coolant conditions are recommended to demonstrate the oxidation rate or lack of 201 one. It is also recommended that in-reactor data from rods with cracked coatings be evaluated 202 to assess if there is aggressive corrosion at cracks or interfaces.

ATF-ISG-01, Appendix B Page 6 of 7 203 B.10: Hydrogen Pickup 204 Zr-alloy Cladding 205 It is important to quantify the hydrogen pickup in uncoated cladding tubes as hydrides in 206 zirconium can lead to brittle behavior of the cladding (Zhao, et al., 2017). Hydrogen from the 207 outer surface is of primary concern as hydrogen from the inner surface is controlled by the fuel 208 fabricators by controls on pellet moisture.

209 Cr-coated Zr 210 In the case of Cr-coated Zr, if it is demonstrated that the metallic or ceramic Cr-coating leads to 211 negligible oxidation and is a barrier to hydrogen pickup, then this might not be necessary for Cr-212 coated Zr cladding tubes. Cracks and defects in the coating may also lead to higher localized 213 hydrogen pickup and lead to cladding damage. Depending on the coating application method, 214 there is potential for hydrogen pickup during coating fabrication. This is expected to be mitigated 215 by process controls.

216 B.11: High Temperature Ballooning Behavior 217 Zr-alloy Cladding 218 The burst stress as a function of temperature is important to know for LOCA analysis as this will 219 determine when to start two-sided oxidation. The ballooning strain is important to determine flow 220 blockage and establish if a coolable geometry has been maintained. Ex-reactor burst tests at 221 temperatures of interest for LOCA on representative cladding segments have been used in the 222 past to establish the high temperature ballooning behavior of Zr-alloy tubes (Powers & Meyer, 223 1980). A significant difference in ballooning behavior between irradiated and unirradiated tubes 224 has not been observed. This is likely due to annealing of radiation defects at burst 225 temperatures.

226 Cr-coated Zr 227 Burst stress and ballooning strain are especially important for Cr-coated cladding as the Cr 228 coating is expected to provide a barrier to high temperature oxidation, but it has not been 229 proposed to coat the inner surface of the tube, so once ballooning and burst has occurred there 230 will be at least some bare Zr available for reaction with high temperature steam. The existing 231 data (see Section 6.2.2) on coated cladding indicate there may be smaller balloon sizes and 232 rupture openings in coated cladding. This may limit high temperature steam on the inner 233 surface. Ex-reactor burst tests at temperatures of interest for LOCA on representative cladding 234 segments would be useful on metallic or ceramic Cr-coated Zr alloy tubes to quantify the 235 ballooning and burst behavior.

236 B:12: High Temperature Steam Oxidation Rate 237 Zr-alloy Cladding 238 The steam oxidation rate is important for LOCA analysis because this determines if the cladding 239 has been overly thinned by corrosion. This also determines the extra heat generation from the 240 corrosion reaction.

241 Cr-coated Zr 242 Ex-reactor oxidation tests at temperatures of interest for LOCA on representative cladding 243 segments have been used to establish the high temperature steam oxidation rate of Zr-alloy

ATF-ISG-01, Appendix B Page 7 of 7 244 tubes. Such data would be useful on either metallic or ceramic Cr-coated Zr alloy tubes to 245 quantify the oxidation rate

1 APPENDIX C 2

3 Specified Acceptable Fuel Design Limits (SAFDLs) 4 5 C.1: SAFDLs Related to Assembly Performance 6 SAFDLs related to assembly performance are typically performed by simple hand calculations 7 or by siting manufacturing controls or historic data. These limits may need revision relative to 8 those typically used for Zr-alloy tubes.

9 C.1.1: Rod Bow 10 Usually there is a penalty on departure from nucleate boiling ratio (DNBR) or margin to critical 11 power ratio (MCPR) to account for bowing. The limits of what degree of bowing is acceptable 12 will not change with the introduction of Cr-coated Zr as this is controlled by the physical 13 dimensions of the fuel assembly. However, bowing methods rely on correlations that are very 14 empirical. Some testing or assessment would be useful to assess the applicability of the rod 15 bow correlation used for Cr-coated cladding. The coating application should result in a uniform 16 thickness as coating non-uniformities could lead to rod bow.

17 C.1.2: Irradiation Growth 18 The assembly design allows for a given amount of growth and will define the limit. The axial 19 growth from Section B.8 will be used to assess maximum growth. There are currently no 20 additional concerns that need to be addressed regarding irradiation growth for Cr-coated Zr 21 cladding.

22 C.1.3: Hydraulic Lift Loads 23 The limits for hydraulic lift loads are such that the upward hydraulic forces do not exceed the 24 weight of the assembly and the downward force of the holddown springs. None of these 25 parameters are expected to change with the introduction of Cr-coated Zr cladding. Existing 26 limits and methods are expected to be adequate.

27 C.1.4: Fuel Assembly Lateral Deflections 28 The limits for fuel assembly lateral deflections are such that the control rod (PWR) or control 29 blades (BWR) can still be inserted as needed. Current assembly and channel bow methods are 30 used to assess performance relative to these limits. Assembly and channel bow are not 31 impacted by fuel rod performance, but rather by channel design (BWR) and guide tube design 32 (PWR) and therefore these limits and methods are not expected to change with the introduction 33 of Cr-coated Zr cladding tubes.

34 C.1.5: Fretting Wear 35 Current design limits state that fuel rod failures will not occur due to fretting. Fretting has 36 historically been controlled though debris filters that reduce the possibility for debris fretting and 37 through spacer design to reduce fretting between fuel rods and grid features. Ex-reactor fretting 38 tests on unirradiated Cr-coated Zr cladding tubes would be useful to ensure that fretting 39 behavior will not be an issue with the coating. A concern for Cr-coated Zr is that grid features 40 are not damaged by the hard coating on the fuel rod. Ex-reactor fretting tests could be used to 41 demonstrate that grids are not damaged by the hard coating on the fuel rod.

ATF-ISG-01, Appendix C Page 2 of 12 42 C.2: SAFDLs Related to Rod Performance Assessed for Normal Operation and AOOs 43 Current codes that are informed by the properties in Section 5.1 can perform the following 44 analyses. However, the limits may need revision relative to those typically used for Zr-alloy 45 tubes. Several of these SAFDLs also have application in accident analysis.

46 C.2.1: Cladding Stress 47 Cladding stress limits are typically set using a method described in Section III of the ASME code 48 (American Scociety of Mechanical Engineers, 2017). Typically, these limits are based on 49 unirradiated yield stress to represent the lowest yield stress. For Cr-coated Zr, the use of the 50 unirradiated yield stress determined in Section A.6 should be acceptable to determine a stress 51 limit.

52 C.2.2: Cladding Strain 53 There are two cladding strain limits that are typically employed. The first steady-state limit is the 54 maximum positive and negative deviation from the unirradiated conditions that the cladding may 55 deform throughout life. The second transient strain limit is the maximum strain increment 56 caused by a transient. This transient cladding strain may also be applicable to accident analysis.

57 These cladding strain limits are typically justified based on mechanical tests (axial tension tests 58 and tube burst tests) performed on irradiated cladding tubes. Ductility tends to decrease with 59 irradiation (Geelhood, Beyer, & Luscher, 2008), so these tests are most relevant when 60 performed at the maximum expected fast neutron fluence. The uniform elongation or strain 61 away from the rupture has been typically used as the strain capability for Zr-based alloys 62 (Geelhood, Beyer, & Cunningham, 2004). This would be a good metric for Cr-coated Zr cladding 63 to protect against cladding mechanical failure. For Cr-coated cladding, there is the additional 64 concern that large strains in the cladding may lead to cracking of the coating (See Section 6.3.1 65 of the PIRT report). Cracking of the coating can lead to a loss of corrosion protection for the 66 substrate along with delamination. It may be desirable to add crack detection criteria so that 67 there is no detectable cracking or microcracking of the coating 68 C.2.3: Cladding Fatigue 69 The cladding fatigue limit is typically based on the sum of the damage fractions from all the 70 expected strain events being less than 1.0. The damage fractions are typically found relative to 71 the ODonnell and Langer irradiated fatigue design curve (O'Donnell & Langer, 1964). It is 72 currently unknown if the ODonnell and Langer irradiated fatigue design curve would be 73 applicable to Cr-coated Zr. It has been noted (Kvedaras, Vilys, Ciuplys, & Ciuplys, 2006) that in 74 steels, Cr coating can improve or significantly worsen the fatigue lifetime due to different 75 microstructures produced in the coating. This was also observed in the case of Cr-coated Zr 76 where the fatigue life went down with the application of a coating (Sevecek, et al., 2018).

77 Because of this, fatigue data from irradiated cladding that was produced using a representative 78 process for the applicant in question is recommended to either confirm the ODonnell and 79 Langer irradiated fatigue design curve or to develop a new fatigue design curve. New fatigue 80 design curves should include a safety factor of 2 on stress amplitude or a safety factor of 20 on 81 the number of cycles as mentioned in the Standard Review Plan Section 4.2.

ATF-ISG-01, Appendix C Page 3 of 12 82 C.2.4: Cladding Oxidation, Hydriding, and CRUD 83 For Zr-alloy cladding, the cladding oxidation limit is designed to preclude oxide spallation that 84 has typically been observed above 100 m. Oxide spallation or coating spallation can lead to a 85 local cool spot which acts as a sink for hydrides, creating a local, extremely brittle hydride lens.

86 The hydrogen limit is designed to ensure that the strain limit previously identified will be 87 applicable since high levels of hydrogen (>600ppm) can cause embrittlement of the cladding.

88 Hydrogen is not the only embrittlement mechanism and there may be other embrittlement 89 mechanisms that are discussed elsewhere. There is no explicit limit on CRUD, other than it be 90 explicitly considered if it is present and it is typically modeled as an insulating layer around the 91 fuel rod in plants that have CRUD issues.

92 None of these limits are particularly relevant to Cr-coated cladding since the outer oxide will be 93 Cr2O3 rather than ZrO2 and the Cr and/or Cr2O3 are expected to be a barrier against hydrogen 94 uptake. Limits should be proposed that preclude environmental damage to the protective Cr2O3 95 layer and embrittlement of the cladding. If intermetallics form on the surface of the cladding, the 96 oxide could be a mixture of ZrO2 and Cr2O3. As with Zr-alloy cladding, the CRUD should be 97 monitored in plants and be explicitly considered if it is present and modeled as an insulating 98 layer around the fuel rod.

99 C.2.5: Fuel Rod Internal Pressure 100 There are several possible limits for rod internal pressure that are discussed in the Standard 101 Review Plan Section 4.2. The first and most straightforward is that the rod internal pressure 102 shall not exceed the coolant system pressure. No outward deformation or hydride reorientation 103 is possible if the stress in the cladding is in the compressive directions. This situation does not 104 change with the application of a Cr coating. Therefore, this limit would still be applicable to Cr-105 coated Zr cladding.

106 Greater rod internal pressures may be justified based on the following criteria:

107

• No cladding liftoff during normal operation 108

• No reorientation of the hydrides in the radial direction in the cladding 109

• A description of any additional failures resulting from departure of nucleate boiling (DNB) 110 caused by fuel rod overpressure during transients and postulated accidents.

111 It has typically been determined by applicants with Zr-alloy cladding that the first of these 112 criteria, no cladding liftoff during normal operation, is the most limiting. This should be confirmed 113 by the applicant of a Cr-coated Zr cladding to still be the case. If this is found to be the case, the 114 pressure limit where cladding liftoff could occur is typically set as the pressure where the upper 115 bound cladding creep rate will exceed the lower bound fuel pellet swelling rate. For Cr-coated Zr 116 cladding, the fuel pellet swelling rate will not be changed and the cladding creep rate will be 117 determined as discussed in Section B.7, provided that the coating does not significantly change 118 the cladding thermal conductivity.

119 C.2.6: Internal Hydriding 120 Internal hydriding is typically addressed through manufacturing controls on the pellet moisture 121 limit. The inner surface for the Cr-coated Zr cladding will be the same and therefore the typical

ATF-ISG-01, Appendix C Page 4 of 12 122 approach would also apply for Cr-coated Zr cladding. It is not expected that the application of a 123 coating will impact this conclusion.

124 C.2.7: Cladding Collapse 125 Cladding collapse in modern nuclear fuel rods has been mitigated by pellet design features such 126 as dishes and chamfers on the ends of the pellet that effectively eliminate axial gaps in the fuel 127 pellet column. Nevertheless, cladding collapse analyses are performed for potential small axial 128 gaps between pellets and in the upper plenum region. The key input into this analysis is the 129 cladding creep rate. For Cr-coated Zr the cladding creep rate will be determined as discussed in 130 Section B.7.

131 C.2.8: Overheating of Fuel Pellets 132 For this analysis, the limit is the melting temperature of the fuel pellets. This will not be impacted 133 by the introduction of Cr-coated Zr cladding and therefore the limit for this SAFDL may stay the 134 same.

135 C.2.9: Pellet-to-Cladding Interaction 136 Typically, there is no explicit limit set on pellet-to-cladding interaction. Various manufacturing 137 designs and inspections and the transient cladding strain limit are expected to cover this 138 SAFDL. The inner surface for the Cr-coated Zr cladding will be the same and therefore the 139 typical approach would also apply for Cr-coated Zr cladding.

140 C.3: SAFDLs Related to Fuel Rod Performance Assessed for Accident Conditions 141 Current codes that are informed by the properties in Appendix A can perform the following 142 analyses. However, the limits may need revision relative to those typically used for Zr-alloy 143 tubes. Several of these SAFDLs also have application in AOO analysis.

144 There is currently work underway to change some regulations (10CFR50.46c) and staff 145 guidance (DG1327) for LOCA and RIA analysis. Neither of these is complete yet, so the 146 discussion in this report will reflect the current regulations and staff guidance.

147 C.3.1: Overheating of the Cladding 148 Overheating of the cladding refers to exceeding the critical heat flux (CHF). This is applicable to 149 AOOs and some accident analyses. Operation above this point results in a reduction of the 150 coolant to remove heat and can result in damage to the cladding. In a PWR, exceeding CHF 151 results in departure from nucleate boiling (DNB). In a BWR, exceeding CHF results in dryout.

152 This thermal margin should not be exceeded for normal operation and AOOs. For design basis 153 accidents the number of fuel rods exceeding thermal margin criteria are assumed to have failed 154 and are included in fission product release dose calculations.

155 The boiling transitions are shown graphically in Figure 5.1 of the PIRT report. Typical limits are 156 based on ex-reactor flow tests on electrically heated fuel assembly mockups to determine where 157 CHF occurs. The CHF is primarily influenced on the geometry of the assembly, although surface 158 conditions of the fuel rods may also impact the CHF. Surface conditions include surface 159 roughness, wettability, and porosity (e.g., of a CRUD layer). Most studies have concluded that 160 roughness has little or no impact on CHF (Collier & Thome, 1994), (Kandlikar, 2001), (O'Hanley,

ATF-ISG-01, Appendix C Page 5 of 12 161 et al., 2013) though some studies have shown a noticeable difference between rough and very 162 smooth surfaces (Weatherford, 1963). Surface porosity and wettability are thought to have a 163 much more significant impact, as demonstrated by several experimental studies (Kandlikar, 164 2001), (Takata, Hidaka, Masuda, & Ito, 2003), (O'Hanley, et al., 2013). Boiling heat transfer 165 experimental results indicate similar CHF for coated and uncoated cladding (Jo, Yeom, 166 Gutierrez, Sridharam, & Corradini, 2018) (Jo, Gutierrez, Yeom, Sridharan, & Corradini, 2019),

167 but given the number or parameters known to impact CHF, it is important to perform CHF tests 168 on each coating and assembly type in question.

169 The application of a coating to fuel rods, while keeping the rest of the assembly the same, is not 170 expected to impact these CHF correlations if the surface conditions of the coating are similar to 171 that of the reference Zr-alloy tubes. It is currently not known what the surface roughness, 172 contact angle, or CRUD deposition rate for a Cr-coated tube will be relative to an uncoated tube.

173 If the coating results in a significantly different surface roughness or cladding outer diameter 174 than the reference Zr-alloy tube, then ex-reactor flow tests on electrically heated fuel assembly 175 mockups with prototypical coated cladding tube could be performed to determine where CHF 176 occurs. Currently, many CHF tests are performed on Inconel assemblies. This may not be 177 appropriate for determining the effect of a coating on Zr cladding.

178 As mentioned in Section 4.1 of the PIRT report, the possibility of formation of a low temperature 179 eutectic between Cr and Zr exists if temperature exceeds 1332°C. This formation should either 180 be considered under this damage mechanism or under generalized cladding melting (Section 181 B.3.7).

182 C.3.2: Excessive Fuel Enthalpy 183 Excessive fuel enthalpy relates to the sudden increase in fuel enthalpy from an RIA below the 184 fuel melting limit that can result in cladding failure due to pellet-cladding mechanical interaction.

185 Current fuel enthalpy limits are based on RIA tests that have been performed on irradiated and 186 unirradiated fuel rodlets in various test reactors and a limit has been determined of what level of 187 fuel enthalpy increase will cause cladding failure.

188 For Zr-alloy cladding, these data have been collected over a very long period and it may not be 189 practical to collect this amount of data for Cr-coated Zr cladding.

190 An alternate approach comes from the fact that cladding failure due to excessive fuel enthalpy is 191 driven by pellet-cladding mechanical interaction which causes the cladding to exceed its ductility 192 limit. Therefore, it is possible to collect uniform elongation (strain at maximum load) data from 193 the irradiated cladding mechanical tests that need to be performed to collect the elastic modulus 194 (Section B.5) and yield stress data (Section B.6). If it can be shown that the Cr-coating has a 195 beneficial or negligible impact on the uniform elongation relative to the reference Zr-alloy 196 cladding, then it could be reasonably argued that the current RIA failure limits are applicable to 197 Cr-coated Zr cladding. If this were the case then a more limited number of RIA tests on Cr-198 coated Zr clad fuel rods may be acceptable, or a commitment to collecting such data could be 199 acceptable.

ATF-ISG-01, Appendix C Page 6 of 12 200 It should be noted that this limit is used to assess the number of fuel rods that are expected to 201 fail during an RIA, and a conservative approach could be taken to either assume all the rods will 202 fail or a significantly conservative limit could be applied to cover the lack of RIA test data on Cr-203 coated Zr cladding.

204 C.3.3: Bursting 205 Bursting of the fuel rod relates to failure of fuel rods due to high temperature and high gas 206 pressures during a LOCA. This can also be a consideration during RIA. It is important to know 207 the rupture stress as a function of temperature and the amount of ballooning that would occur.

208 There are no specific design limits associated with cladding rupture other than that the degree 209 of swelling not be underestimated and the balloon not block the coolant channel. Additionally, 210 the time of rupture needs to be known so that oxidation on the cladding inner surface and its 211 associated heat is correctly modeled.

212 An applicant will typically use an empirical correlation for burst stress and ballooning strain such 213 as the one given in NUREG-0630 (Powers & Meyer, 1980). If an applicant uses NUREG-0630 214 for Cr-coated Zr cladding, it would be useful to collect some data to show that the performance 215 of Cr-coated Zr is bounded by these limits. Alternatively, if the applicant wants to propose new 216 burst stress and ballooning strain limits, a significant body of burst data would be useful to 217 demonstrate that the degree of swelling not be underestimated. Currently available data 218 suggest that for Cr-coated cladding, the balloon region is smaller and burst temperature 219 increases (see Section 6.2.2 of the PIRT Report), however, this should be confirmed for the 220 specific coating in question.

221 C.3.4: Mechanical Fracturing 222 Mechanical fracturing refers to a defect in the cladding caused by an externally applied force.

223 Typically, this limit has conservatively been set as applied stresses above 90% of the irradiated 224 yield stress. This limit should not be exceeded for normal operation and AOOs. For design basis 225 accidents the number of fuel rods exceeding this limit are assumed to have failed and are 226 included in fission product release dose calculations.

227 This limit is acceptable for Cr-coated Zr cladding given that the irradiated yield stress obtained 228 as described in Section B.6 is used.

229 C.3.5: Cladding Embrittlement 230 Cladding embrittlement relates to embrittlement of the fuel cladding, particularly in the ballooned 231 region of the cladding during LOCA. Cladding embrittlement during LOCA should be precluded 232 so the fuel assemblies with ballooned rods are not severely damaged by post LOCA loads such 233 as reflood and quenching, including blowdown loads. 10 CFR 50.46 specifies a cladding 234 temperature limit of 2200°F 5.19 (1204°C) and a peak oxidation of 17% equivalent cladding 235 reacted for Zr-alloy cladding (US Nuclear Regulatory Commission, 2017).

236 The PIRT ranked this damage mechanism as high. (See Appendix A of the PIRT report). It is 237 not known if these limits will be acceptable for Cr-coated Zr cladding. It appears as if the outer 238 surface will reduce the high temperature metal-water reactor from that of bare Zr, but it is 239 unknown if some other mechanism could cause embrittlement of the cladding. One possible

ATF-ISG-01, Appendix C Page 7 of 12 240 mechanism could be Zr-Cr interdiffusion as discussed in Section 4.2 of the PIRT report. The 241 formation of a brittle rim of ZrCr2 could lead to brittle cladding failure similar to how the formation 242 of a dense hydride rim can lead to brittle cladding failure.

243 Tests showing ductility (See Section 6.2.6 of the PIRT report) at either these existing limits or 244 test establishing new limits would be useful to demonstrate embrittlement will not occur. In 245 addition to the tests performed to establish the ballooning (Section B.11) and high temperature 246 oxidation behavior (Section B.12), some prototypic integral LOCA tests (see for example 247 (Flanagan, Askeljung, & Puranen, 2013)) where cladding tubes are subject to ballooning and 248 burst in steam under expected time frames and samples are then subjected to mechanical 249 loading such as bend tests after ballooning, burst, and high temperature oxidation are very 250 useful to establish cladding embrittlement limits. For these tests, irradiated cladding tubes are 251 preferable.

252 C.3.6: Violent Expulsion of Fuel 253 Violent expulsion of fuel relates to the sudden increase in fuel enthalpy from an RIA that can 254 result in melting, fragmentation, and dispersal of fuel. This could result in a loss of coolable 255 geometry and produce a pressure pulse that could damage the reactor vessel. Typical limits for 256 violent expulsion of fuel are:

257

• Peak radial average fuel enthalpy below 230 cal/g 258

• Peak fuel temperature below melting temperature.

259 It is expected that cladding failure will occur well before 230 cal/g for both Zr-alloy and Cr-260 coated Zr cladding. These limits are derived to prevent violent ejection of fuel from failed 261 cladding. As such, these limits relate more to the fuel than to the cladding and are expected to 262 be appropriate for Cr-coated Zr cladding.

263 C.3.7: Generalized Cladding Melting 264 Generalized cladding melting is applicable to design basis accidents and is set to preclude the 265 loss of coolable geometry. The limit is set as the cladding melting temperature, which for Zr is 266 1852°C. For Zr alloy tubes the embrittlement limit of 1204°C (Section C.3.5) is more limiting.

267 However, as discussed in Section B.3.5, it is unknown what the limit for Cr-coated Zr 268 embrittlement will be, so cladding melting should still be considered for Cr-coated Zr.

269 The melting temperature of Cr (1857°C) is virtually identical to that of Zr (1852°C). However, the 270 formation of a low temperature eutectic between Cr and Zr at 1332°C occurs significantly lower 271 than either of the individual melting temperatures. Formation of a low temperature eutectic with 272 a thin coating may not represent loss of geometry such as generalized cladding melting, but the 273 formation of the eutectic should either be considered under this damage mechanism or under 274 overheating of the cladding (Section C.3.1).

275 C.3.8: Fuel Rod Ballooning 276 Ballooning of the fuel rod relates to failure of fuel rods due to high temperature and high gas 277 pressures during a LOCA. It is important to know the rupture stress as a function of temperature 278 and the amount of ballooning that would occur. There are no specific design limits associated

ATF-ISG-01, Appendix C Page 8 of 12 279 with cladding rupture other than the degree of swelling not be underestimated and the balloon 280 not block the coolant channel.

281 An applicant will typically use an empirical correlation for burst stress and ballooning strain such 282 as the one given in NUREG-0630 (Powers & Meyer, 1980). If an applicant uses NUREG-0630 283 for Cr-coated Zr cladding, it would be useful to collect some data to show that the performance 284 of Cr-coated Zr is bounded by these limits. Alternatively, if the applicant wants to propose new 285 burst stress and ballooning strain limits, a significant body of burst data from either unirradiated 286 or irradiated cladding tubes would be useful to demonstrate that the degree of swelling not be 287 underestimated.

288 C.3.9: Structural Deformation 289 Structural deformation refers to externally applied loads during LOCA or safe shutdown 290 earthquake that could deform the fuel assemblies or cause fuel fragmentation such that 291 coolable geometry would be lost. This limit has conservatively been set as applied stresses 292 above 90% of the irradiated yield stress. For design basis accidents the number of fuel rods 293 exceeding this limit are assumed to have failed and are included in fission product release dose 294 calculations.

295 This limit is acceptable for Cr-coated Zr cladding given that the irradiated yield stress obtained 296 as described in Section A.6 is used.

297 C.4: New Damage Mechanisms 298 There have been several new damage mechanisms identified for Cr-coated Zr cladding. These 299 may either be addressed by applicants through existing limits or as separate limits. The 300 following sections identify those new damage mechanisms that have been identified for Cr-301 coated Zr through a technical review of the recent data and a general understanding of coating 302 behavior. Each section will identify the potential for fuel system damage, fuel rod failure, or 303 impact on fuel coolability. These sections will also identify existing SAFDLs that could be used 304 to account for these damage mechanisms. These damage mechanisms are physical 305 mechanisms and should be addressed even if no credit for coating performance is credited in 306 the fuel system safety review.

307 C.4.1: Coating Cracking 308 Cracking of the coating could occur during the relatively large (0.5% to 1% strain) deformations 309 that are observed occur in the cladding due to cladding thermal expansion, cladding creepdown, 310 deformation of the cladding due to pellet swelling, and axial irradiation growth. Cracking could 311 also occur in the cladding due to repeated small strain (0.01% to 0.1% strain) cyclic operation.

312 Finally, cracking could occur during a design basis accident that causes large strain from pellet 313 expansion (RIA) or gas overpressure and ballooning (LOCA).

314 The PIRT ranked this damage mechanism as high during accident conditions. (See Appendix A 315 of the PIRT report). Excessive cracking of the coating could eliminate the benefit that the 316 coating provides for normal operation (reduced in-reactor corrosion and hydrogen pickup) as 317 well as during accident conditions (may expose significant amount of Zr to high temperature 318 steam). Cracking of the coating could also create crack tips that extend into the Zr cladding that

ATF-ISG-01, Appendix C Page 9 of 12 319 could provide stress concentrations for further environmentally assisted crack mechanisms and 320 could ultimately lead to cladding failure.

321 Cracking of the coating should be considered in the development of the cladding strain limit 322 (Section C.2.2) and the cladding fatigue limit (Section C.2.3). In these cases, it should be 323 considered if failure is defined when cracking of the coating is observed. Cracking of the coating 324 should also be considered in the development of high temperature ballooning (Section B.11) 325 and high temperature oxidation (Section B.12) correlations. If cracking is observed following 326 ballooning, then high temperature oxidation correlations should be developed based on cladding 327 with a cracked coating. Additionally, cladding embrittlement limits (Section C.3.5) should be 328 developed based on cracked cladding.

329 C.4.2: Coating Delamination 330 Delamination of the coating could occur due to a variety of reasons including poor adherence to 331 the substrate and differential thermal expansion between the coating and the substrate. In 332 general, ceramic coatings will be more susceptible to delamination than metallic coatings.

333 The PIRT ranked this damage mechanism as high during accident conditions. (See Appendix A 334 of the PIRT report). Delamination of the coating could eliminate the benefit that the coating 335 provides for normal operation (reduced in-reactor corrosion and hydrogen pickup) as well as 336 during accident conditions (may expose significant amount of Zr to high temperature steam) 337 depending on the amount of delamination. Local coating delamination could create a local cool 338 spot on the cladding which is a sink for hydrogen diffusion. This local cool spot could develop a 339 hydride blister that results in local brittle cladding behavior. Finally, coating delamination can 340 increase the quantity of debris in the reactor coolant system which could lead to enhanced 341 debris fretting and could impact the performance of emergency core coolant system pump in the 342 event of an accident if the debris filters become clogged with debris from delaminated coating.

343 Debris clogging this pump has been identified as Generic Safety Issue 191 (GSI-191) (Shaffer, 344 et al., 2005).

345 Delamination of the coating should be considered in the development of the cladding strain limit 346 (Section C.2.2) and the cladding fatigue limit (Section C.2.3). In these cases, it should be 347 considered if failure is defined to be observed delamination of the coating. Delamination of the 348 coating should also be considered in the development of high temperature ballooning (Section 349 B.11) and high temperature oxidation (Section B.12) correlations. If delamination is observed 350 following ballooning, then high temperature oxidation correlations should be developed based 351 on cladding with a delaminated coating. As discussed in Section 4.2, the ZrCr2 phase that could 352 form due to interdiffusion could exhibit greater corrosion rate than bare Zr. Additionally, if this is 353 the case, cladding embrittlement limits (Section C.3.5) should be developed based on 354 delaminated cladding. LOCA blowdown loads could also lead to delamination of the coating. To 355 address GSI-191, the potential for delamination should be evaluated and accounted for 356 following burst (Section C.3.3), mechanical fracture (Section C.3.4), ballooning (Section C.3.8),

357 and structural deformation (Section C.3.9).

ATF-ISG-01, Appendix C Page 10 of 12 358 C.4.3: Cr-Zr Interdiffusion 359 As discussed in Section 4.2, if temperatures at the Cr-Zr interface and the time at temperature 360 are great enough there will be the formation of a CrZr intermetallic that is more brittle than either 361 Cr or Zr separately. If this intermetallic layer is thick enough, it could lead to brittle cladding 362 failure. Thin layers of this intermetallic would likely not reduce the overall cladding ductility.

363 However, the critical thickness for overall brittle behavior is not known. The calculations from 364 Section 4.2 are shown below.

365

• Normal Conditions (300°C-350°C for 2000 days) 0.1 to 0.3 m thick intermetallic layer 366

• Loss-of-coolant Conditions (800 to 1200°C for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) 0.2 to 1.4 m thick intermetallic 367 layer 368

• Long term Loss-of Coolant (800 to 1200°C for 1 day) 1 to 7 m thick intermetallic layer.

369 Initial data from a number of programs has not observed significant interdiffusion in various 370 coating concepts. It is noted that the numbers above are predictions based on limited data and 371 should not be used without any data from a coating in question.

372 Unless otherwise accounted for in specific strain or ballooning limits, the formation of this CrZr 373 intermetallic should be avoided. During normal operations and AOOs, the temperature at the 374 Cr/Zr interface is only expected to allow for the formation of a very thin CrZr intermetallic layer, 375 but during design basis accidents the cladding temperature may be large enough to form a 376 significant thickness of this layer (See Section 4.2 of the PIRT report). Other possibilities for the 377 formation of the CrZr intermetallic phase include during application of the coating if the substrate 378 temperature is too great, and during the welding of end caps in the heat affected zone of the 379 weld.

380 The Cr/Zr intermetallic is both brittle and exhibits extremely poor high temperature corrosion 381 behavior (See Section 4.2 of the PIRT report). If a significant thickness of Cr/Zr intermetallic 382 were to form during high temperature conditions during a design basis accident or some 383 manufacturing process, the cladding could behave in a brittle manner, the corrosion reaction 384 may worsen, and various design limits on strain and cladding embrittlement may no longer be 385 applicable.

386 Cr-Zr interdiffusion should be considered in the development of limits on overheating of the 387 cladding (Section C.3.1), clad embrittlement (Section C.3.5), and eutectic formation related to 388 generalized clad melting (Section C.3.7). If some Cr-Zr interdiffusion is caused during the 389 manufacturing process, then it should be ensured that limits are developed on prototypic parts 390 from this process and tests are performed in localized areas known to have the possibility for 391 interdiffusion.

392 C.4.4: Radiation Effects on Cr 393 It has been noted that the irradiation of Cr will result in the formation of the radioisotope Cr-51 394 with a half-life of 28 days. It is known that this isotope will be formed, but it is not known if this 395 isotope will be released to the coolant in significant quantities. For a CrN coating, the nitrogen 396 will lead to the production of some C-14. A second concern is what the impact of fast neutron 397 irradiation on Cr metal and other Cr containing compounds will be. In zirconium, fast neutron

ATF-ISG-01, Appendix C Page 11 of 12 398 irradiation leads to a dramatic increase in strength and reduction in ductility (Geelhood, Beyer, &

399 Luscher, 2008). Recent ion beam irradiation data indicated that cold spray Cr-coatings are more 400 resistant to radiation defects than bulk Cr. (Maier B. , et al., 2018) 401 The release of Cr-51 from the cladding into the coolant could challenge the plant dose release 402 limit or the ability of the chemical and volume control system to eliminate Cr ions before they 403 plate out on the fuel and the other reactor components. The impact of fast neutron irradiation on 404 the strength and ductility of the Cr metal or other Cr containing compounds could lead to a 405 degradation in coating performance beyond what we expected based on tests on unirradiated 406 material.

407 The formation and possible release of Cr-51 is an issue that may be monitored through ongoing 408 surveillance at the plant. Plants already have a process in place to evaluate the radioisotopes 409 and the gaseous and liquid effluents and report this information to the NRC on an annual basis.

410 If Cr-51 in the coolant begins to challenge plant dose release limits, it will be observed to 411 increase as more of the fuel in the core is transitioned to Cr-coated Zr cladding. In this case, 412 systems can be implemented to effectively remove this radioisotope before it becomes a safety 413 problem. Similarly, with the impact of Cr ions on the coolant chemistry, a surveillance plan put in 414 place alongside the implementation of Cr-coated Zr cladding to monitor the coolant chemistry 415 will mitigate any impact of Cr ions. The impact of fast neutron irradiation on Cr mechanical 416 properties will be inherently included in material property correlations and limits that are 417 developed based on irradiated material as described in previous sections.

418 C.4.5: Subsurface Damage 419 As mentioned in Section 3.0 of the PIRT report, many physically bonded coating systems may 420 require mechanical preparation such as grit blasting to obtain a suitable surface for coating 421 bonding. It is currently unknown what the impact of this surface preparation will be on the 422 performance of the coated cladding. The impact will undoubtedly be highly process dependent 423 and should be evaluated for each qualified coating in question.

424 C.4.6: Residual Stress 425 When coatings are applied at a different temperature than their application temperature, it is 426 possible to develop residual stress in the cladding and the coating. This stress could lead to 427 unexpected cladding or coating failure. It is currently unknown what the impact of this residual 428 stress will be on the performance of the coated cladding. The impact will undoubtedly be highly 429 process dependent and should be evaluated for each qualified coating in question.

430 C.4.7: Galvanic Corrosion 431 Galvanic corrosion refers to corrosion damage induced when two dissimilar materials are 432 coupled in a corrosive electrolyte. It occurs when two (or more) dissimilar metals are brought 433 into electrical contact under water. Galvanic corrosion can be accelerated under the effects of 434 radiation as has been observed with the so-called shadow corrosion observed between BWR 435 channel boxes and control blades. When a galvanic couple forms, one of the metals in the 436 couple becomes the anode and corrodes faster than it would all by itself, while the other 437 becomes the cathode and corrodes slower than it would alone.

ATF-ISG-01, Appendix C Page 12 of 12 438 Dissimilar metals in this case, include: Cr+Zr, Inconel+Cr, and CrN+Zr. No indication of galvanic 439 corrosion, irradiation assisted or otherwise between these systems has been found in this effort.

440 LTA data may be used to further clarify if this will be a problem.

441 C.4.8: Defects 442 Any coating process will result in some population of defects. Depending on the size and 443 concentration of these defects, they could lead to oxidation under the coating either in normal 444 operating conditions or accident conditions. This could lead to cracking or delamination of the 445 coating which could eliminate the benefits of the coating and have other safety consequences 446 (see Sections C.4.1 and C.4.2). The PIRT ranked this damage mechanism as high during 447 accident conditions. (See Appendix A of the PIRT report). Each process in question should 448 define the allowable defects and justify the presence of these defects based on testing of 449 cladding with similar defect concentrations.

450 C.4.9: Eutectic Formation 451 The formation of eutectics seems to be a concern primarily for beyond design basis accident 452 conditions. The lowest temperature eutectic for the Cr-Zr system occurs at 1332°C. If operation 453 beyond the current design basis temperature limit of 1200°C is requested, then the formation of 454 eutectics and their impact on the coating should be considered. Additionally, in systems other 455 than the Cr-Zr system, such as Cr-Zr-N, the formation of lower temperature eutectics should be 456 considered for both design basis and beyond design basis accident conditions.