ML19198A306

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

ML19198A306
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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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DRAFT INTERIM STAFF GUIDANCE 1

SUPPLEMENTAL GUIDANCE REGARDING THE 2

CHROMIUM-COATED ZIRCONIUM ALLOY FUEL CLADDING 3

ACCIDENT TOLERANT FUEL CONCEPT 4

ATF-ISG-01 5

6 7

PURPOSE 8

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

14 15 BACKGROUND 16 17 This interim staff guidance (ISG) is intended to provide guidance for NRC staff reviewing 18 applications involving fuel products with chromium-coated zirconium alloy cladding. For coated 19 claddings of this type, a phenomena identification and ranking table (PIRT) was generated for 20 the NRC by Pacific Northwest National Laboratory; the guidance provided in this ISG 21 extensively references the PIRT report, Degradation and Failure Phenomena of Accident 22 Tolerant Fuel Concepts: Chromium Coated Zirconium Alloy Cladding (Reference 1). The 23 suggested cladding properties, specified acceptable fuel design limits (SAFDLs), and new 24 failure mechanisms sections from the PIRT are replicated in Appendix B and C, so that 25 modifications to the information may be made based on stakeholder comments and feedback.

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

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

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

38

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

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

52 53 APPLICABILITY 54 55 This guidance applies to:

56 57 All holders of an operating license or construction permit for a nuclear power reactor under Title 58 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic Licensing of Production and 59 Utilization Facilities, except those who have permanently ceased operations and have certified 60 that fuel has been permanently removed from the reactor vessel.

61 62 All holders of and applicants for a power reactor early site permit, combined license, standard 63 design approval, or manufacturing license under 10 CFR Part 52, Licenses, Certifications, and 64 Approvals for Nuclear Power Plants. All applicants for a standard design certification, including 65 such applicants after initial issuance of a design certification rule.

66 67 All holders of and applicants for a power reactor early site permit (ESP), combined license 68 (COL), standard design certification (DC), standard design approval (DA), or manufacturing 69 license (ML) referencing a small modular reactor (SMR) design under Title 10 of the Code of 70 Federal Regulations (10 CFR) Part 52, Licenses, Certifications, and Approvals for Nuclear 71 Power Plants. SMRs are defined using the International Atomic Energy Agency definition of 72 small and medium-sized reactors with an electrical output of less than 700 megawatts.

73 74 All contractors and vendors (C/Vs) that supply basic components to U.S. Nuclear Regulatory 75 Commission (NRC) licensees under Title 10 of the Code of Federal Regulations (10 CFR) Part 76 50, Domestic Licensing of Production and Utilization Facilities or 10 CFR Part 52, Licenses, 77 Certifications, and Approvals for Nuclear Power Plants.

78 79

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

86 87 IMPLEMENTATION 88 89 The staff will use the information contained in this ISG to ensure that all known degradation and 90 failure mechanisms for chromium-coated zirconium alloy fuel cladding are considered such that 91 their impact on the acceptance criteria contained in SRP sections 4.2, 4.3, and 4.4 along with 92 chapter 15 can be assessed.

93 94 BACKFITTING AND ISSUE FINALITY DISCUSSION 95 96 Discussion to be provided in final ISG.

97 98 CONGRESSIONAL REVIEW ACT 99 100 Discussion to be provided in final ISG.

101 102 FINAL RESOLUTION 103 104 By 2025, this information will be transitioned into Chapters 4 and 15 of the SRP. Following the 105 transition of this guidance to the SRP, this ISG will be closed.

106 107 108 APPENDICES 109 110 A.

Supplemental Guidance for SRP Chapters 4 and 15 111 B.

Cladding Material Property Correlations 112 C.

Specified Acceptable Fuel Design Limits (SAFDLs) 113 D.

(Placeholder) Resolution of Public Comments 114 115 REFERENCES 116 117

1. Chromium-Coated Cladding Final PIRT Report, Degradation and Failure Phenomena of 118 Accident Tolerant Fuel Concepts: Chromium Coated Zirconium Alloy Cladding, June 119 2019 (Agencywide Documents Access and Management System (ADAMS) Accession 120 No. ML19172A154) 121

ATF-ISG-01 Page 4 of 4

2. NRCs Standard Review Plan, Standard Review Plan for the Review of Safety Analysis 122 Reports for Nuclear Power Plants: LWR Edition (NUREG-0800), (ADAMS Accession 123 No. ML070810350) 124

3. NRCs ATF Project Plan, Project Plan to Prepare the U.S. Nuclear Regulatory 125 Commission for Efficient and Effective Licensing of Accident Tolerant Fuels, September 126 2019 (ADAMS Accession No. ML18261A414) 127 128 Public Meetings: August 6, 2019; December 4, 2019 129 130

APPENDIX A 1

2 Supplemental Guidance for SRP Chapters 4 and 15 3

4 NUREG-0800 - Chapter 4, Section 4.2, Fuel System Design 5

For reviews of new fuel products where the only change from an existing approved fuel design 6

that utilizes zirconium alloy cladding is the adoption of chromium-coated cladding, the licensing 7

of a new cladding alloy can be used as a model. While SRP 4.2 covers additional requirements 8

for review of complete fuel systems, cladding reviews cover these three areas:

9 Definition of specified acceptable fuel design limits (SAFDLs) for new cladding, 10 Material property correlations to be used in codes to ensure the new cladding satisfies 11 the SAFDLs, and 12 Any changes that must be made to existing methodologies to accommodate the new 13 cladding.

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

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

17 Harder surface 18 o Improves cladding fretting performance and wear resistance 19 20 Negligible oxidation during normal operation 21 o Protects zirconium cladding from oxidation 22 o Protects zirconium cladding from hydrogen uptake 23 24 Improved high temperature steam oxidation kinetics 25 o Reduced rate of corrosion and heat of oxidation 26 o Protects zirconium cladding from oxidation 27 o Reduced hydrogen liberation 28 29 Improved high temperature strength 30 This ISG does not attempt to set standards for review of any credit or benefit applicants may 31 request by demonstrating these improvements, as strategies for licensing these potential 32 improvements have not yet been submitted to the Nuclear Regulatory Commission. The 33 reviewer of any coated cladding must, therefore, evaluate any proposed property improvements 34 against the data provided by the applicant. The reviewer must also evaluate if the data provided 35 supports the full operating domain for the fuel, and place appropriate limitations and conditions 36 when necessary. Finally, if an applicant wishes to take credit for coating behavior up to a certain 37 burnup, or during certain accident conditions, it is necessary for the adherence of that coating to 38 the substrate to have been demonstrated to that burnup and during those conditions.

39

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

42 SAFDLs related to fuel assembly performance that are typically addressed by simple 43 calculation, manufacturing controls, and historical data 44 SAFDLs related to fuel rod performance that are typically addressed for normal 45 operation and anticipated operational occurrences (AOOs) using a thermal mechanical 46 code 47 SAFDLs related to fuel rod performance that are typically addressed for accident 48 conditions using a system analysis code with initial conditions provided by a thermal 49 mechanical code.

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

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

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

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

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

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

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

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

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

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

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

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

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

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

117

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

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

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

131 Fuel temperature 132 Fission gas release 133 Rod internal pressure and void volume 134 Cladding oxide thickness 135 Cladding permanent hoop strain following a power ramp.

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

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

139 Identification of functional requirements for the fuel and assembly 140 Identification of limits for each functional requirement 141 Identification of code or other approach that will be used to assess performance against 142 functional requirement 143 Identification of approach to demonstrate high level of confidence that design will not 144 exceed functional requirements:

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

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

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

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

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

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

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

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

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

184

1) Changes to hydraulic diameter due to the coating thickness 185

2) Changes to boiling crisis behavior, including effects of surface roughness 186

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

268 269

APPENDIX B 1

2 Cladding Material Property Correlations 3

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

6 thermal conductivity 7

thermal expansion 8

emissivity 9

enthalpy and specific heat 10 elastic modulus 11 yield stress 12 thermal and irradiation creep rate (function of stress, temperature, and fast neutron flux) 13 axial irradiation growth 14 oxidation rate 15 hydrogen pickup.

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

20 High temperature ballooning behavior 21 High temperature (800-1200°C) steam oxidation rate.

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

161 162

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

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

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

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

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

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

202

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

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

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

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

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

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

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

APPENDIX C 1

2 Specified Acceptable Fuel Design Limits (SAFDLs) 3 4

C.1: SAFDLs Related to Assembly Performance 5

SAFDLs related to assembly performance are typically performed by simple hand calculations 6

or by siting manufacturing controls or historic data. These limits may need revision relative to 7

those typically used for Zr-alloy tubes.

8 C.1.1: Rod Bow 9

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

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

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

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

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

41

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

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

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

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

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

81

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

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

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

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

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

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

106 No cladding liftoff during normal operation 107 No reorientation of the hydrides in the radial direction in the cladding 108 A description of any additional failures resulting from departure of nucleate boiling (DNB) 109 caused by fuel rod overpressure during transients and postulated accidents.

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

199

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

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

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

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

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

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

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

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

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

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

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

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

256 Peak radial average fuel enthalpy below 230 cal/g 257 Peak fuel temperature below melting temperature.

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

356 and structural deformation (Section C.3.9).

357

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

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

364 Normal Conditions (300°C-350°C for 2000 days) 0.1 to 0.3 m thick intermetallic layer 365 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 366 layer 367 Long term Loss-of Coolant (800 to 1200°C for 1 day) 1 to 7 m thick intermetallic layer.

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

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

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

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

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

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

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

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

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

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

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

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

437

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

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

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

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

456