The use of small graphite specimen test data for large core components for HTGR

1. The use of small graphite specimen test data for large core components for HTGR Presented at the ASTM Symposium on “Graphite Testing for Nuclear Applications: Significance of Test Specimen Volume and Geometry and the Statistical Significance of Test Specimen Population" Seattle, Washington, September 19-20 (2013). Dr. Makuteswara Srinivasan, Senior Materials Engineer Division of Engineering, Office of Nuclear Regulatory Research US Nuclear Regulatory Commission Washington, DC 20555 The views presented here are those of the author only and do not necessarily represent those of the US Nuclear Regulatory Commission. This presentation was prepared, in part, by the author, who is an employee of the U.S. Nuclear Regulatory Commission on his own time apart from his or her regular duties. NRC has neither approved nor disapproved its technical content.

2. Guidance Provided for Presentation • Personal perspective on how the regulatory agency (NRC) views and uses ASTM testing methods in their assessments of license approval • how does NRC deal in general with the data from nonstandard tests - Is there an overarching rule-of-thumb or guideline that a regulatory agency normally uses for data from nonstandard tests? • same thing for small population samples. This is especially pertinent to the NRC perspective since they have to deal with very little irradiated data whether if is for graphite or metals or whatever • Discussion about any NRC rule or regulation that deals directly with this specific issue would be very helpful • Personal perspective on small specimens and small sample populations based upon your years at the NRC but also from your other experiences in materials Address these kinds of questions which would give the testers and researchers a perspective from the regulatory side of things

3. Presentation Plan • Background – NRC and Its Regulatory Role • The Use of Codes and Standards – Guidance Documents • Light water reactor (LWR) Application Examples • Metals – Reactor Pressure Vessel (RPV) and Piping • Concrete • Other Materials • LWR Technical Issue Examples • Small sample test experience • Use of simulated environment in laboratory tests • Use of “non-standardized” tests • Inadequate test procedure • Need for standard test

4. Presentation Plan • Specific Research Needs for HTGR to Better Understand and Communicate – Technical Basis for Developing Regulatory Guidance Documents: • Standardization of Material Test Reactor (MTR) Data And Inclusion of Their Limitations – Use of “Non-Standard” Sizes and Geometry • Data Dispersion For Large Graphite Core Components (Non-irradiated) Using ASTM Standards • Methods Used to Extrapolate MTR Data - Consistency And Acceptable Methods • Uncertainties in Data, Assumptions, And Model • Verification And Validation Demonstration Using Round Robin Testing on Non-Irradiated Graphite • Adoption Of Recommended Practice and Use in ASME Codes and Standards Development

5. Mission of the NRC The Nuclear Regulatory Commission regulates the civilian uses of nuclear materials in the United States to protect public health and safety, the environment, and the common defense and security. The mission is accomplished through licensing of nuclear facilities which possess, use and dispose nuclear materials; the development and implementation of requirements governing licensed activities; and inspection and enforcement activities to assure compliance with these requirements.

6. Regulatory Responsibilities 10 CFR Parts 1 -199 Nuclear Regulatory Commission Staff Reviews Licensee’sSafety Analysis Report (SAR)Using Standard Review Plan (SRP) Staff Uses Guidance Documents, such as Regulatory Guides (RG),codes and standards (C&S),and industry documents,for guidance on acceptable procedures,methodology for technical review

7. Guidance Documents • Regulatory Guide (RG) • Provides information on staff position on technical issues and acceptable techniques or methods to address them and meet regulations in license application and staff evaluation • The methods outlined in RG are not mandatory; the licensee may propose and use alternate methods and solutions by providing basis for complying with appropriate regulations • Interim Staff Guidance (ISG) • Resolve emerging regulatory or inspection issues

8. Regulatory Requirements for Components General Design Criteria (GDC) 1, 2, 4 and 10 and 10 CFR Part 50, §50.55a require that structures and components important to safety shall be constructed and tested to quality standards commensurate with the importance of the safety functions to be performed (GDC 1), and designed with appropriate margins to withstand effects of anticipated operational occurrences (GDC 10), normal plant operation; natural phenomena such as earthquakes (GDC 2); postulated accidents including loss-of-coolant accidents (LOCA), and from events and conditions outside the nuclear power unit (GDC 4).

9. Regulatory Requirements for Components The application of GDC 1 requirement to the reactor internals provides assurance that established standard design practices of proven or demonstrated effectiveness are used to achieve a high likelihood that these safety functions will be performed. The application of GDC 2 to the reactor internals provides assurance that they will withstand earthquakes combined with the effects of normal or accident conditions. The application of GDC 4 to the reactor internals provides assurance that the effects of environmental conditions to which they are exposed over their installed life will not diminish the likelihood of performance of these safety functions under all operating conditions, including accidents. This provides assurance that failures of the reactor internals resulting from environmental service conditions that could cause loss of capability to monitor reactivity, fuel damage resulting from loss of reactivity control, structural damage to fuel cladding, or interference with core cooling are not likely to occur.

10. Regulatory Requirements for Components The application of GDC 10 to the reactor internals provides assurance that they are designed with sufficient margin to ensure their functionality and integrity during any condition of normal operation, including the effects of anticipated operational occurrences, such that a high likelihood of performance of these safety functions is achieved. Assured performance of these safety functions in turn assures that specified acceptable fuel design limits related to reactivity control and core cooling are not exceeded, thus assuring the integrity of the fuel and its cladding.

11. Regulatory Review of Component Design • The regulatory staff reviews, among other information: • The physical or design arrangements of all reactor internals structures, components, assemblies, and systems, including the manner of positioning and securing such items within the reactor pressure vessel, the manner of providing for axial and lateral retention and support of the internals assemblies and components, and the manner of accommodating dimensional changes due to thermal and other effects. • The loading conditions that provide the basis for the design of the reactor internals to sustain normal operation, anticipated operational occurrences, postulated accidents, and seismic events. All combinations of listed design and service loadings (e.g., operating pressure differences and thermal effects, seismic loads, and transient pressure loads associated with postulated loss-of-coolant accidents) that are accounted for in design of the reactor internals.

12. Regulatory Review of Component Design • The regulatory staff reviews, among other information: • The design bases for the mechanical design of the reactor vessel internals, including allowable limits such as maximum allowable stresses; stability under dynamic loads; deflection, cycling, and fatigue limits; and core mechanical and thermal restraints (positioning and hold-down). • Each combination of design and service loadings, categorized with respect to the allowable design or service limits (defined in the ASME Code), and the stipulated associated stress intensity or deformation limits. Design or service loadings should include safe shutdown earthquake (SSE) and operating basis earthquake (OBE) loads as appropriate.

13. Discussion of Light Water Reactor Experience Examples

14. Regulatory Use of Codes and Standards • Codes and standards are incorporated into the rule in 10 CFR 50.55a “Codes and Standards • The unacceptable Code Cases and the reasons for non-acceptance are provided in RG 1.193. • insufficiency in technical basis • errors in equations, • inaccurate configuration mapping to real components • uncertainties in implementation, such as lack of performance demonstration • ASME Section III Code Cases not yet endorsed by the NRC may be used by a licensee or applicant through 10 CFR 50.55a(a)(3). • This section permits the use of alternatives to the Code requirements referenced in 10 CFR 50.55a provided that the proposed alternatives result in an acceptable level of quality and safety, and that their use is authorized by the Director of the Office of Nuclear Reactor Regulation

15. Regulatory Use ofIndustry Documents • The NRC staff evaluates industry guidance documents, such as those from Nuclear Energy Institute (NEI) and Electric Power Research Institute (EPRI) and their use by licensees in safety evaluation • NEI-03-08, “Guideline for the Management of Materials Issues” • EPRI BWRVIP and PWRVIP documents • EPRI MRP documents • Where applicable, cited in Regulatory Guides • The NRC staff evaluates documents from PWR Owners Group (PWROG) and BWR Owners Group (BWROG) • Provide specific experimental and/or methodology to address both generic and plant-specific emerging and already encountered technical issues and plans to address them

16. Examples of Key Metals Properties for Safety Evaluation Reactor environment (water chemistry, temperature, load (stress), loading rate and cumulative neutron fluence) and material condition (different materials, different heats, different fabrication methods (such as forging and rolling), different heat treatments (mill-annealed and solution-annealed) influence the following properties: Yield Strength Strain-to-failure Fracture Toughness Corrosion Resistance Fatigue Resistance Propensity for Crack Initiation Propensity for Crack Propagation

17. Boric Acid Corrosion in PWR Environment Several variables can influence the extent of corrosion, which may depend on the testing variable and material, thus exhibiting wide variations in both the amount of material loss due to corrosion and the rate of corrosion. (Non-standard) Testing carried out under simulations of potential plant conditions: • immersion of carbon steel in de-aerated borated water • immersion of carbon steel in aerated borated water • dripping of borated water onto hot carbon steel surfaces • impingement of borated water onto hot carbon steel surfaces • leakage of borated water into a tight crevice (nozzle applications) Significant variability in corrosion rate data of approximately 6 orders of magnitude was observed, demonstrating the test variables, such as nature of the test and test temperature on the corrosion of various carbon and low alloy steels Ref: E.S. Hunt “Boric Acid Corrosion Guidebook, Revision 1: Managing Boric Acid Corrosion Issues at PWR Power Stations”, EPRI, Palo Alto, CA: 2001. 1000975

18. Boric Acid Corrosion in PWR Environment NRC sponsored research at Argonne National Laboratory - used an ASTM G5-94 standard test procedure. ANL conducted in three distinct environmental conditions to understand the corrosion performance of reactor steels exposed to H-B-O system. • room-temperature-saturated boric acid solution in aerated and de-aerated conditions for electrochemical potential (ECP) and corrosion experiments at 95 °C • molten-salt solutions with and without water additions for corrosion tests at 150-320 °C and 1 atmosphere • high-pressure (12.4 MPa) high-temperature (up to 316 °C) water with a range of boric acid solution concentrations for ECP and corrosion experiments. • Significant corrosion was observed only for the low–alloy steel and no corrosion for Alloy 600 or 308 stainless steel cladding. Also, corrosion rates were significant in saturated boric acid solutions, and no material loss was observed in H-B-O solution in the absence of moisture. Ref: J.-H. Park, O. K. Chopra, K. Natesan, and W. J. Shack, “Boric Acid Corrosion of Light Water Reactor Pressure Vessel Materials”, NUREG/CR-6875, U.S. Nuclear Regulatory Commission, July 2005, NRC ADAMS Accession No: ML052360563

19. Embrittlement of Reactor Pressure Vessel (RPV) An example of testing of a representative specimen to assess the performance of a reactor component • Objective: Protect RPV against pressurized thermal shock events • Applicable Rule: • 10 CFR 50.61 requires maintenance of adequate fracture toughness for the RPV. • 10 CFR 50, Appendix A, General Design Criterion (GDC) 31 requires that the design reflect the uncertainties in determining the effects of irradiation on material properties. • Appendix O, "Fracture Toughness Requirements," and Appendix H, "Reactor Vessel Material Surveillance Program Requirements," which implement, in part, Criterion 31, necessitate the calculation of changes in fracture toughness of reactor vessel materials caused by neutron radiation throughout the service life

20. Guidance For Meeting the PTS Rule Regulatory Guide 1.161, “Evaluation of Reactor Pressure Vessels With Charpy Upper-Shelf Energy Less Than 50 ft-lb”, June (1995). • It is recognized that the specific material of interest (i.e., the material from the beltline region of the reactor vessel under operation) is seldom available for testing. Thus, testing programs have used generic materials that are expected to represent the range of actual materials used in fabricating reactor pressure vessels (RPV) in the United States. • The NRC has accepted statistical analyses of such generic data for use in the methods described in this guide. • The RG also states that other methods for determining the material property may be used on an individual-case basis, if justified.

21. Guidance For Meeting the PTS Rule Regulatory Guide 1.99, Revision 2, “Radiation Embrittlement of Reactor Vessel Materials”, U.S. Nuclear Regulatory Commission, May (1988) • describes general procedures acceptable to the NRC staff for calculating the effects of neutron radiation embrittlement of the low-alloy steels currently used for light-water-cooled reactor vessels • two measures of radiation embrittlement are acceptable from the results of the Charpy V-notch impact test • requires that a full curve of absorbed energy versus temperature be obtained through the ductile-to-brittle transition temperature region • The second measure of radiation embrittlement is the decrease in the Charpy upper-shelf energy level, which is defined in ASTM B 185-82. Thus, the NRC has developed acceptable procedures recognizing the practical difficulties in obtaining significant representative irradiation data to assess vessel integrity over time.

22. Use of Embrittlement Trend Curves (ETC) • Embrittlement trend curves (ETC) are used to estimate the magnitude of neutron irradiation embrittlement as a function of: • environmental variables, such as fluence, flux, and temperature • materials composition variables, such as copper, nickel, and manganese • Worldwide data from material test reactors (MTR) with fast fluence, high neutron dose exposures with surveillance coupon data from power reactors consisting of relatively lower cumulative neutron dose, consisting of over 2,500 data have been analyzed. • An empirical algebraic relationship, termed as WR-C(5) model, has been derived; this model comprehensively addresses the basic influences of flux, fluence, and contents of copper, manganese, phosphorus, nickel, and modification terms to consider variations in composition, irradiation environment, and component manufacture. • Influence of the data and model uncertainties on the estimated shift in temperature in the Charpy curve at the 30-foot-pound energy level has been analyzed. Ref: Kirk, Mark, “A Wide-Range Embrittlement Trend Curve for Western Reactor Pressure Vessel Steels,” Effects of Radiation on Nuclear Materials: 25th Volume on June 15–17, 2011 in Anaheim, CA; STP 1547, Takuya Yamamoto, Guest Editor, pp. 20–51, doi:10.1520/STP103999, ASTM International, West Conshohocken, PA 2012

23. Fatigue in Primary (Class 1) Components • An example of the difficulty in obtaining test data for design under specific operating environment is mechanical fatigue of reactor components. • Most LWR fatigue data have been from experiments conducted in air. The variables that can affect fatigue life in air and LWR environments were broadly classified into three groups of material, loading, and environment. • The component design has used ASME Section III Code fatigue curves, which were obtained from strain–controlled fatigue tests of small polished specimens at room temperature in air. Best–fit curves to the experimental test data were first adjusted to account for the effects of mean stress and then lowered by a factor of 2 on stress and 20 on cycles (whichever was more conservative) to obtain the design fatigue curves. • These factors were not safety margins but rather ‘adjustment factors’ to experimental data to account for the effects of material variability and data scatter, as well as size, surface finish, and loading history in low cycle fatigue to obtain estimates components useful life.

24. Fatigue in Primary (Class 1) Components NRC sponsored research at Argonne National Laboratory to understand environmental effects on fatigue. • Results have shown that: • Under certain environmental and loading conditions, fatigue lives in water relative to those in air can be a factor of: • ≈12 lower for austenitic stainless steels • ≈ 3 lower for Ni-Cr-Fe alloys • ≈ 17 lower for carbon and low-alloy steels • A consideration of thermal loading due to flow stratification or mixing can also affect the fatigue behavior. • This work has suggested revising the ASME Code using various (fatigue penalty) factors, Fen, specific for different materials, to incorporate the environmental effects in calculating the fatigue cumulative usage factor (CUF) Ref: “Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials (NUREG/CR-6909) - Final Report”, Available from Web-link: http://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr6909/

25. Effect of Multi-Variables in Modeling It is common that many variables, which often influence one another during reactor operation, influence a key property and affects the outcome of material behavior. • One way to handle such effect is the use of “ratio”, which is the property determined at ambient conditions to that determined at the environment . Fen is such a ratio (using water at service temperature). • Concept of ‘transformed variable’: ANL developed statistical fatigue life relationships to incorporate the variability in material (such as sulfur), environment (such as dissolved oxygen, temperature), and loading (for example, strain rate) using the concept from the best fit of the experimental data • Concept of ‘modification terms’: Was used for the case of predictive models for embrittlement of RPV, where variations in composition, irradiation environment, and component manufacture were treated by developing statistical correlations of experimental data

26. Flaw Evaluation Analysis • The ASME flaw evaluation procedures for nuclear pressure vessels are contained in ASME Section XI, Appendix A. • ASME Section XI specifies periodic volumetric examinations of primary system pressure-retaining components that are generally accomplished using ultrasonic (UT) examination techniques. • To simplify the analysis of indications detected by such examinations, Section XI adopted the conservative assumption that all observed indications are planar (crack-like) defects. • Elastic-plastic fracture mechanics (EPFM), linear elastic fracture mechanics (LEFM), and limit load calculations are then performed on the determined flaw to estimate the stress intensity factor (K) of this flaw to the critical stress intensity factor (KIc). • In conducting safety evaluation, the NRC staff assesses licensee’s calculations to ensure that the evaluation conservatively predicted an end-of-life size of the indication, using credible loading and degradation mechanisms. In order for an indication to be considered acceptable by the ASME Code, Section XI, the indication must be calculated to remain smaller than 75 percent of the through-wall thickness for axial flaws.

27. Difficulties encountered in component inspection • Component nondestructive inspection methods are usually developed and used based on laboratory test experience, component mock-up test experience, and operating experience. • An event at North Anna Power Station Unit 1, on March 24, 2012. • Two through-wall cracks were discovered after machining the Unit 1 'B' Steam Generator (SG) Hot Leg Nozzle to provide a flat surface for welding and to provide a configuration that would better facilitate future inspections. • The licensee concluded that the postulated flaw depth exceeded the limits of applicability in the code of a/t=0.75 and that the as-found flaws, while potentially within acceptable stress limits, did not meet the ASME Section XI code requirements. • The direct cause determined the leakage was due to primary water stress corrosion cracking (PWSCC) in a susceptible material (i.e. Alloy 82/182). • A combination of tensile stresses and low chromium weld material on the inner diameter of the dissimilar metal weld in a high temperature, primary water environment directly led to the formation of stress corrosion cracks Ref: “LER 12-001-00 for North Anna, Unit 1, Regarding Degraded Reactor Coolant System Piping Due To Primary Water Stress Corrosion Cracking”, NRC ADAMS Accession No: ML12151A441. May (2012).

28. Discussion on Concrete

29. Concrete in Nuclear Power Plant • The concrete containment is one of the most important components of an NPP. It functions as the final barrier to the release of fission products to the outside environment and is independent of the fuel barrier and reactor coolant pressure boundary barrier. • Other concrete structures include: • containment internal structures • secondary containment • reactor buildings • fuel and equipment storage pools • emergency cooling water structures Ref: C. J. Hookham, In-Service Inspection Guidelines for Concrete Structures in Nuclear Power Plants, ORNL/NRC/LTR-95/14, Oak Ridge National Laboratory, Oak Ridge, Tennessee, 1995

30. Degradation of Concrete • Similar to graphite, concrete can undergo a volume change when exposed to radiation due to the behavior of the concrete aggregate. Fast neutrons displace atoms, which results in considerable growth as measured in certain aggregate (e.g., flint). • In addition to damage caused due to gamma heating, long-term exposure to irradiation can decrease in the tensile and compressive strengths and modulus of elasticity of concrete.

31. Degradation of Concrete • A phenomenon of considerable concern is the alkali-silicate chemical reaction (ASR). • The primary factors influencing alkali-silica reactions include the aggregate reactivity (i.e., amount and grain size of reactive aggregate), alkali and calcium concentrations in water used for concrete content (i.e., alkali content), presence of water, and temperature. • An ASTM method exists to test concrete for susceptibility to the ASR. This standard also requires confirmation by supplemental information by petrographic examination of the concrete prisms to identify the products of alkali-silica reactivity using Practice C856. • However, this laboratory test has been criticized as not adequately representing field conditions. • The accuracy of these relatively short-term reactive aggregate tests and the potential for sudden occurrence after long-term performance have been identified as technical issues Ref: (1) Special NRC Oversight at Seabrook Nuclear Power Plant: Concrete Degradation”. http://www.nrc.gov/info-finder/reactor/seabrook/concrete-degradation.html (2) K. Ono, “Damaged Concrete Structures in Japan Due to Alkali Silica Reaction,” International Journal of Cement Composites and Lightweight Concrete 10(4), 247 (1988).

32. Testing Concrete Strength • The effects of size and shape on the strength and other properties of concrete. • Typically, cube (U.K. and Germany) and cylinder (Australia, Canada, France, New Zealand and the United States) shapes have been used for testing concrete. • Research conducted by the NY Department of Transportation has indicated the preference for the use of cylinders. However, debate continues in the technical community to better understand the merits of each of these shapes and sizes. • Test sizes depend on the aggregate size. It is usually agreed, a “rule of thumb”, that the diameter of the cylindrical specimens to the smallest dimension of a rectangular cross section must be at least three times the nominal maximum size of coarse aggregate used in concrete. • It has been postulated that there also exists differences in the type of test, such as flexural vs. tensile. Elwet, D. J., & Fu, G. “Compression Testing of Concrete: Cylinders vs. Cubes.” Special Report 119, Transportation Research and Development Bureau, New York State Department of Transportation, March (1995)

33. Examples of Other Materials

34. Testing of Charcoalfor Air Filtration RG 1.52, “Design, Inspection, and testing Criteria for Air Filtration and Adsorption Units of Post-Accident Engineered-Safety-Feature Atmosphere Cleanup Systems in Light-Water-Cooled Nuclear Power Plants” • Several ASME standards and one ASTM standard have been stated to be acceptable to NRC staff for design and testing of engineered safety feature atmosphere cleanup systems • Invokes IAEA Safety Guide, consistent with NRC’s interest in facilitating the harmonization of standards used domestically and internationally • Also mentions an IEEE Standard for designing all associated instrumentation and controls • The NRC staff has also identified instances where technical issues were identified in laboratory testing where inappropriate testing conditions not representative of filed conditions were used and the results obtained not conservative. • The use of outdated test protocols or inappropriate test conditions could potentially overestimate the charcoal’s ability to adsorb radioiodine following an accident Ref: 1. U.S Nuclear Regulatory Commission; “Deficiencies in the Testing of Nuclear-Grade Activated Charcoal;” Information Notice 87-32; Washington, DC 20555; (1987). 2. U.S. Nuclear Regulatory Commission; “Laboratory Testing of Nuclear-Grade Activated Charcoal;” Federal Register: Vol. 63, No. 37; pages 9581-9589; Washington, DC 20555; (1998), See also: NRC Generic Letter 99-02, (1999), See: http://www.nrc.gov/reading-rm/doc-collections/gen-comm/gen-letters/1999/gl99002.html

35. Testing of Neutron Absorbers • The NRC Generic Letter, GL-96-04 provides detailed information on the degradation of Boraflex in spent fuel pools. • Although an ASTM Standard exists for qualification of such material for use in dry cask storage, standards need to be developed and are currently underway for use in wet systems. • A standard testing procedure has not yet been developed is Boraflex, which is used in spent fuel racks as a neutron absorber. Ref: NRC Generic Letter 96-04, “Boraflex Degradation in Spent Fuel Pool Storage”, See: http://www.nrc.gov/reading-rm/doc-collections/gen-comm/gen-letters/1996/gl96004.html.

36. ASME Design Code For HTGR Graphite Core Components • Published in the Section III New Division 5 (2013 edition) • Requires component design to account for the effects of irradiation on the thermal and mechanical properties • Statistical variation of strength within the billet and variations in properties, for different production runs are to be incorporated in the design • Mandatory appendix HHA-I of the Code has accepted ASTM material specification standards for graphites • ASME Code Article HHA-III-3000 also states that the irradiated properties data shall be reported in accordance with ASTM C625 • Code allows test procedures other than standard tests, provided such procedures are identified with the materials test data • Article HHA-III-4200 of the ASME Code recognizes the difficulties in obtaining test data from representative samples and test standards and assigns the responsibility to the graphite core designer for the determination and justification of the representative data.

37. Status of NRC’s Review ofC&S for HTGR NRC is yet to review : ASME Code for the design of graphite core components in HTGR All ASTM standards relevant to graphite (HTGR) NRC has no current plans to initiate activities related to developing staff guidance documents for potential review of HTGR design A key need is to generate and assemble key properties information data, predictive models, and generally accepted interpretation protocols for such information for design and safety evaluation of graphite core components. Such information will provide the necessary technical basis for NRC staff evaluation and development of staff positions on technical issues in meeting the requirements of the reactor regulation and GDC, and develop Regulatory Guides specific to graphite core components for HTGR

38. Graphite Component Integrity Graphite Component Integrity Physical Properties Thermal Properties Mechanical Properties Nuclear Properties Density ExpansionContractionConductance Young’s ModulusPoisson’s RatioStrength and Strength DistributionFracture ResistanceCreep and Fatigue LimitOxidation ResistanceWear and Erosion Resistance Cross-Section Need Correlation† And Understanding OfNon-Irradiated AndIrradiated Properties Graphite Microstructure RawMaterials PorosityPore Size and Shape DistributionPore Orientation and DistributionGrain Size and Shape DistributionGrain Orientation and Distribution ManufacturingMethod Normal OperationsAccidents Part Sizeand Geometry † Correlation does not imply cause; PIE and analysis may shed light.

39. Inspection Model Graphite Degradation Model Risk Model ComponentIntegrityModel Specific PRA Tools for Graphite Components GraphiteComponent

40. Classes of Irradiation Data (Model) Uncertainty • Epistemic: lack-of knowledge uncertainties arising because our scientific understanding of irradiation data is imperfect for the present, but are of a character that in principle are reducible through further research and gathering of more and better irradiation data. • Aleatory - "random" (stochastic) in character; uncertainties in models which for all practical purposes cannot be known in detail or cannot be reduced. • Even under "perfect information“, i.e., when the model has been validated and the numerical values of its parameters are known, these aleatory uncertainties are still present (for a given model). • Approximations (predictions deviate by a fixed but unknown amount from observed values of the predicted value). • uncertainties about the numerical values of the parameters of a given model Adapted from: Ref: R. J. Budnitz G. Apostolakis, D. M. Boore, L. S. Cluff, K. J. Coppersmith, C. A. Cornell, and P. A. Morris “Recommendations for Probabilistic Seismic Hazard Analysis: Guidance on Uncertainty and Use of Experts”, NUREG/CR-6372 (1997).

41. Development Needs for Using Irradiation Data for Codes and Standards • Development of procedures to interpret MTR irradiation data, consistent with ASTM properties determination specifications • Development of procedures to interpolate and extrapolate limited MTR irradiation data, quantifying the uncertainties in the data and model • Development of methods to check for consistency in irradiation data of “non-standard” specimen size and geometry • Validation and verification of the properties behavior model • Formulation of consensus standard procedure to use MTR irradiation data for large specimen sizes (components) • Incorporation of the standard in ASME graphite design codes

42. Considerations for Next Step • Demonstrate proof-of-concept using non-irradiated nuclear graphite specimens • Fabricate specimens conforming to: • Existing ASTM standard size and geometry for various significant physical and mechanical properties • Several selected size and geometry being used for current NGNP irradiation programs • Determine various physical and mechanical properties • Compare and interpret data for consistency and conformance to material microstructural sampling and representation, including quantification of uncertainties • Develop procedures to extrapolate information from MTR data to properties prediction for large graphite blocks with typical non-homogeneity in microstructure • Recommend range of allowable specimen sizes and geometry for future graphite irradiation.

43. Summary • The LWR industry and the U.S. and other international regulators are aware and are experienced in assessing issues related to data obtained from representative samples of components in simulated environments to actual reactor operation conditions. • Logistic, operational, and economic considerations limit the size and geometry of material test reactor (MTR) specimens or surveillance coupons to deviate from recommended ASTM standard sizes and geometry • Influence of multi-variables in experimental conditions affecting property response, including potential interactions between variables • Influence of materials variables affecting property response, including potential interactions between variables • Technical issues in assessing component response to reactor operational conditions and limited, small sample experimental data, including modeling uncertainty • Technical issues related to predictions of behavior and assessment of component condition from online monitoring and periodic inspection data • Importance of operating experience, including deviations in operating variables, informing predictive model refinement

44. Summary • For HTGR, enabling research to establish technical bases and inform regulatory guidance development may include: • Recommended procedures to relate (non-standard) MTR properties to those of ASTM standard sizes and geometry, where applicable • Consensus methods to interpolate and extrapolate irradiation data to “large” graphite component sizes, considering the epistemic and aleatory uncertainties, including their quantification • A round-robin (non-irradiated) properties determination exercise to examine the consistency and conformance of physical and mechanical properties data, obtained from non-standard MTR-type specimen size and geometry to current ASTM specimen size and geometry enabling consensus standards for irradiation testing and data interpretation.

45. List of Abbreviations

46. List of Abbreviations

47. Contact: makuteswara.srinivasan@nrc.gov Web Link: www.nrc.gov