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DOE OSTI2013

Results from tight and loose coupled multiphysics in nuclear fuels performance simulations using BISON

The behavior of nuclear fuel in the reactor environment is affected by multiple physics, most notably heat conduction and solid mechanics, which can have a strong influence on each other. To provide credible solutions, a fuel performance simulation code must have the ability to obtain solutions for each of the physics, including coupling between them. Solution strategies for solving systems of coupled equations can be categorized as loosely-coupled, where the individual physics are solved separately, keeping the solutions for the other physics fixed at each iteration, or tightly coupled, where the nonlinear solver simultaneously drives down the residual for each physics, taking into account the coupling between the physics in each nonlinear iteration. In this paper, we compare the performance of loosely and tightly coupled solution algorithms for thermomechanical problems involving coupled thermal and mechanical contact, which is a primary source of interdependence between thermal and mechanical solutions in fuel performance models. The results indicate that loosely-coupled simulations require significantly more nonlinear iterations, and may lead to convergence trouble when the thermal conductivity of the gap is too small. We also apply the tightly coupled solution strategy to a nuclear fuel simulation of an experiment in a test reactor. Studying the results from these simulations indicates that perhaps convergence for either approach may be problem dependent, i.e., there may be problems for which a loose coupled approach converges, where tightly coupled won't converge and vice versa. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Projection Analysis for ATR Irradiation of the AFC-FAST Experiment

Analyses of the Advanced Fuels Campaign Fission Accelerated Steady-state Test (AFC-FAST) in the Advanced Test Reactor are presented. A detailed methodology was employed to better account for uncertainties in the planned power and duration of sequential reactor loading cycles. By performing coupled depletion analyses at multiple power levels and durations, the differences in experiment heating outputs can be found. The effects of these uncertainties upon multiple experiment configurations were assessed in an effort to streamline the process of planning for and documenting future irradiations. In conclusion, the data generated from this work have been used to help inform assumptions on subsequent projections to perform only a nominal case depletion.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Analysis of Radiological Release From Fueled Irradiation Experiments During Manual Handling (Slides)

Irradiation experiments are manually handled at the Advanced Test Reactor (ATR) by qualified operators using long handled tools. Standard handling evolutions include insertion and removal from the reactor vessel, transfer to/from a storage location or cask in the ATR canal, and handling/reconfigurations at a canal working tray. Such routine handling has the potential to result in mechanical damage to the experiment boundary which is credited to retain fission products. Since damage can occur due to operator error, this is an anticipated occurrence. This work determines the radiological consequences to receptors inside the ATR facility, as well as public receptors. Given the wide variety of fuel types tested at the ATR, a generic approach to the analysis is taken. The radiological inventory is determined to bound a variety of fuel types (e.g., ceramic and molten fuel matrices) and fission powers on the U-235 enrichment spectrum. The source term analysis considers different release fractions to bound different fuel types and burnups. The postulated handling events occur underwater; thus, retention of the released isotopic content is considered within the canal water column. Retention of radionuclides in the water column is determined using a modified approach of United States (US) Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.183. Radiological dose to the facility receptor is determined using a compartment release model. Dose to the public receptor is determined using atmospheric dispersion models using site specific atmospheric conditions with the use of the Radiological Safety Analysis Computer (RSAC) program version 7.2.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗

DOE OSTI2019

Innovative Implementation of Consolidated Dry Fuel Storage at the INL - 19655

An innovative nuclear-fuel storage bucket design provides increased fuel storage capacity at the Irradiated Fuel Storage Facility (IFSF) on the Idaho National Laboratory (INL) Site. The IFSF is a dry storage facility that is at approximately 93% of its fuel storage capacity. This innovative fuel storage bucket provides for higher density packaging, thereby allowing for the receipt of more fuel at the IFSF and subsequent completion of the wet-to-dry fuel transfer scope at the INL's Idaho Nuclear Technology and Engineering Center (INTEC). The subject fuel originates from the Advanced Test Reactor (ATR) at the INL Site. All spent nuclear fuel (SNF) at INTEC is required by the Idaho Settlement Agreement (ISA) to be out of wet storage by December 31, 2023. The ISA is a tri-party agreement between the U.S. Department of Energy, the U.S. Navy, and the State of Idaho. Currently, the INTEC CPP-666 pools contain all of the INL's SNF remaining in wet storage, and INTEC's Spent Fuel Project is on schedule to meet the ISA deadline. Interestingly, the INL Site's environmental cleanup contract awarded to Fluor Idaho in 2016 also initially required the Spent Fuel Project to continue SNF receipts from the ATR into wet storage at the CPP-666 pools. The continued receipts of ATR SNF into the CPP-666 basin were able to be redirected to the IFSF because of the new storage package design, enabling the Spent Fuel Project to meet the ISA requirement to have all SNF out of wet storage by 2023. Complicating the need to redirect ATR receipts from wet storage is the lack of interim storage capacity at the ATR site. This lack of ATR storage capacity threatened to terminate reactor operations if the SNF could not be received at INTEC. The contractual requirement for continued ATR SNF receipts to prevent termination of reactor operations is resolved by the implementation of the innovative consolidated dry storage configuration in the IFSF. Consolidated fuel storage at the IFSF provides increased fuel storage capacity to facilitate meeting the ISA and the Fluor Idaho contract requirements. It provides for the completion of the ATR wet-to-dry fuel transfer scope and supports continued operation of the ATR by allowing the receipt of additional ATR SNF directly into IFSF dry storage. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2018

Graphite Specimen Mass Measurements for NESHAP Calculations

The Advanced Graphite Creep (AGC) experiments were designed to irradiate various types of graphite specimens to gain an understanding of how the physical and thermal properties of the graphite change. The graphite specimens are machined from nuclear grade graphite that is very pure relative to other graphite types. However, it still contains parts per million (PPM) levels of metal contaminants that are dispersed throughout it. During irradiation in the Advanced Test Reactor (ATR) these metal contaminants become activated resulting in the graphite specimens being classified as a radioactive material. Following their irradiation, a variety of measurements and experiments are performed on these specimens that require them to be heated above 100°C.Whenever a radioactive material is heated to above 100°C the National Emission Standards for Hazardous Air Pollutants (NESHAP) calculations require the activity content, in Curies, of the isotopes that could potentially be released to the environment. The purpose of this ECAR is to document the method by which the mass loss of each sample is determined and how the isotopic content of this mass loss is determined.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Progress on Pu-238 Production at Idaho National Laboratory From March 2021 to February 2022

Idaho National Laboratory has completed irradiation of Np-237 targets in the Advanced Test Reactor’s (ATR) South Flux Trap (SFT) and I-7 positions. INL also progressed qualification of new ATR Gen 1 Np-237 targets for the North East Flux Trap (NEFT), inner A, and H positions. This paper gives an overview of operational and technical activities from March 2021 to February 2022.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2020

Preliminary Evaluation of Loading DOE Standard Canisters in the INL CPP-603 Irradiated Fuel Storage Facility - 20543

This paper looks at the equipment and operations necessary to load United States Department of Energy (DOE)-owned Spent Nuclear Fuel (SNF) into DOE Standard Canisters in the CPP-603 Irradiated Fuel Storage Facility (IFSF) in the Idaho Nuclear Technology and Engineering Center (INTEC) area at Idaho National Laboratory (INL). Two types of fuels are looked at in this evaluation: Advanced Test Reactor (ATR) fuel (uranium-aluminide fuel with aluminum cladding) and Peach Bottom fuel (thorium-uranium carbide fuel in a graphite matrix). The fuel ready for loading would come from fuel storage canisters in the CPP-603 facility. The paper describes the facility, the fuel types, the DOE Standard Canisters, and existing equipment; lists the needed loading operations; reviews facility features and equipment to perform the operations; and then lists the decisions, analyses, designs, demonstrations, and modifications that will be needed to perform the loading of DOE-owned SNF into DOE Standard Canisters in the CPP-603 IFSF. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2024

FUEL PERFORMANCE SIMULATION OF HIGH BURNUP FUELS IN PLANNED INTEGRAL DESIGN BASIS ACCIDENT EXPERIMENTS

High burnup (HBu) fuel rods from the Byron Nuclear Generating Station (BNGS) were recently received at Idaho National Laboratory (INL) to support a variety of planned Nuclear Energy fuel cycle R&D objectives ranging from fuel performance, fuel recycle, and spent fuel research topics. Among these R&D activities, these fuel rods will be the subjects of multiple in-pile experiment programs at the Transient Reactor Test (TREAT) facility as well as detailed characterization and testing in the hot cells at INL and Oak Ridge National Laboratory (ORNL). TREAT RIA experiments are planned for the Nuclear Energy Agency Framework for Irradiation Experiments (FIDES) Joint Experimental Program called High burnup Experiments in Reactivity Initiated Accident (HERA) program. TREAT and ORNL-furnace LOCA experiments are part of the Department of Energy (DOE) Advanced Fuels Campaign (AFC) program U.S. consensus LOCA test plan, and the in-pile experiments have also been proposed in a FIDES project called Loss of Coolant-High Burnup (LOC-HBu). The results of these test programs will provide crucial data about safety performance enabling extended licensable burnup limits for these fuels. The purpose of this paper is to document fuel performance computational simulations of the BNGS fuel using the Bison code. The detailed assessments include (1) the irradiation history of the fuel to provide prediction of as-run fuel conditions and (2) extending the irradiated fuel conditions into the as-designed experiment conditions for the HERA-HBu RIA experiments and for the LOC-HBu LOCA experiments. The results of these assessments will inform post-irradiation examinations (PIE) of the BNGS parent rods and detailed final design of the planned experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

FUEL PERFORMANCE SIMULATION OF HIGH BURNUP FUELS IN PLANNED INTEGRAL DESIGN BASIS ACCIDENT EXPERIMENTS

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Progress on Pu-238 Production at INL From March 2021 to February 2022

Idaho National Laboratory has completed irradiation of Np-237 targets in the Advanced Test Reactor’s (ATR) South Flux Trap (SFT) and I-7 positions. INL also progressed qualification of new ATR Gen 1 Np-237 targets for the North East Flux Trap (NEFT), inner A, and H positions. This slide show is based upon artifact INL/CON-22-65729-Rev000 and gives an overview of operational and technical activities from March 2021 to February 2022.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2022

Effect of Water Chemistry on Crack Growth Rates in Neutron Irradiated X-750 and XM-19

The Advanced Test Reactor (ATR) National Science User Facility (NSUF) was used to study the effect of neutron damage on crack growth rates (CGR) under simulated boiling water reactor (BWR) water chemistry conditions using a fast neutron spectrum to accelerate the damage process. Nickel based Alloy X-750 and XM-19 were irradiated in the ATR center flux trap (CFT), with this experiment marking the first civilian project to utilize this position within the reactor. 0.4T-CT specimens were used to study crack growth rates for two sample fluences (~1.9 x 1020 n/cm2 and ~9.5 x 1020 n/cm2, E > 1MeV) in both normal water chemistry (NWC) at 2.5 ppm O2 and hydrogen water chemistry (HWC) at ~80 ppb H2. A decrease in CGR was observed with HWC for both alloys with the average ratio (CGR in NWC/CGR in HWC) being a factor of 281 for the XM-19 samples and a factor of 27.8 for the X-750 samples. Observed CGRs in this study showed little difference compared to non-irradiated samples (the non-irradiated XM-19 was 20% cold worked), suggesting limited sensitivity of these susceptible alloys to irradiation effects. The CGR response to varying environmental electrochemical potential (ECP) and water purity will be presented.

36 MATERIALS SCIENCE↗

DOE OSTI2022

Neutronic Safety Analysis of Pu-238 Production at Idaho National Laboratory

This analysis was completed to support the irradiation of plutonium fuel services (PFS) targets in the NEFT in the Advanced Test Reactor (ATR) as a part of the campaign to restart domestic production of plutonium-238 used in radioisotope power systems (RPS) by the National Aeronautical and Space Administration (NASA) and Department of Energy (DOE) Office of Nuclear Energy (NE), Office of Nuclear Infrastructure Program (NE-3). Referred to as the PFS-ATR-GEN1-NEFT experiment, the assembly was designed to hold 46 PFS targets in the NEFT. The scope of this paper is to outline the MOPY method used to calculate the heat generation rates (HGRs), flux, fission density, and quantify the viability of the target design for Pu-238 production in ATR.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2016

A Radial Heat Flow Apparatus for Thermal Conductivity Characterisation of Cylindrical Samples

During the last two decades, silicon carbide based ceramic composites (SiC/SiC) have become a candidate material for nuclear fuel cladding, first for advanced nuclear systems, such as the GFR, and recently also as potential replacement material for nuclear fuel cladding in light water reactors. This work is contributing to the General Atomics/Westinghouse led CARAT project. In this frame, the thermal conductivity (TC) of cladding prototype sections are measured both before and after neutron irradiation with the purpose of better understanding the effects of neutron irradiation on the pyrolytic carbon interphase linking the SiC fibres to the SiC matrix. To this end, an apparatus has been designed and built to measure the TC of short cylindrical samples, whose lengths range from 5 to 20 cm, by the radial heat flow method. To avoid oxidation issues, the experiments are carried out in vacuum. Today the most common measurement method used to characterise the TC of materials is the so-called laser-flash technique, whereby a short high intensity laser pulse is shot on one surface of a thin flat sample and the energy transfer through the sample is analysed on the other side. Though much faster and easier to implement than the apparatus presented in this work, laser-flash equipment does have issues when dealing with highly anisotropic materials or capturing shape effects such as those present in large aspect ratio samples such as cladding tube sections. The two aforementioned effects are exactly what the developed equipment has been built to capture. The radial heat flow method is based on the principle of measuring the temperature difference building up through a piece of material as this material acts as a thermal resistance against a known heat flow crossing it. This paper presents a new implementation of one of the classical thermal conductivity measurement techniques. Since the development of the laser-flash method in the sixties, steady-state measurements such as the radial heat flow have fallen in disuse. In this discussion, it is shown that a direct measurement method can be used and even be preferable in the case of highly anisotropic materials such as SiC/SiC composites. The advantage of this experimental setup is that the actual transverse conductivity is obtained, whereas laser-flash measurements yield a response in which both the transverse and axial conductivities are mixed. Further steps of this study will include the measurements of samples irradiated with neutrons in the MITR test-reactor. The aim will be to evidence the role played by the pyrolytic carbon interphase by linking the TC measurements with energy filtered electron microscopy. We would like to acknowledge General Atomics and Westinghouse for their support in the frame of the CARAT research program. Providing us with actual prototype clad sections is a unique opportunity. Furthermore the Swiss CCEM.ch project MeAWaT is acknowledged for partially financing the development of the measurement device. (authors)

36 MATERIALS SCIENCE↗

DOE OSTI2014

AGR-2 Daily As-run Thermal Analyses

This report documents the daily as-run thermal analyses performed on Advanced Gas-cooled Reactor (AGR) experiment AGR-2 in the Advanced Test Reactor (ATR) for capsules 2,3,5, and 6. Capsule 1 (French) results are in Appendix F, while Capsule 4 (South African) results are in Appendix G. Appendix F, and G are separate documents Several fuel and material irradiation experiments, which support the development of the Next Generation Nuclear Plant (NGNP), are planned for the Advanced Gas Reactor Fuel Development and Qualification Program. The goals of these experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post-irradiation examination (PIE) and safety testing. AGR-2 was the second in this series of planned experiments to test tristructural-isotropic (TRISO)-coated, low enriched uranium oxycarbide (LEUCO) fuel. The AGR-2 experiment was intended to serve as a follow-on to the AGR-1 experiment. The AGR-2 experiment is a high temperature fuel particle test. The test train planned for AGR-2 is based on the experience gained from previous irradiations in the ATR, using instrumented lead experiments. Instrumented lead experiments are used for irradiations requiring a controlled environment and monitored parameters. The experiment test train positions the fuel within the test location and contains sweep gas lines and thermocouple wiring that is routed through access ports to external support systems. The AGR-2 experiment is comprised of six individual capsules, approximately 1.375-in. in diameter by 6-in. long, stacked on top of each other to form the test train. Each capsule contains 12 fueled compacts that are approximately 0.5-in. in diameter by 1-in. long. The compacts are composed of fuel particles bound together by a carbon matrix. Each compact contains approximately 4,150 fissile particles (35 vol% particle packing fraction). Each capsule will be supplied with a flowing helium/neon gas mixture to control the test temperature and to sweep any fission gases that are released to the fission product monitoring system. Temperature control is accomplished by adjusting the gas mixture ratio of the two gases (helium and neon) with differing thermal conductivities. Many similarities exist between this thermal analysis for AGR-2 with the AGR-1 daily as-run report in [1]. Gas gaps were taken from as built drawings in Reference [2]. The finite element heat transfer code ABAQUS [3] was used to perform the thermal analysis. The AGR-2 experiment was placed in the B-12 position in the ATR core as shown in Figure 1. The AGR-2 experiment capsule assembly consists of six capsules axially stacked. Each capsule contains a graphite holder with three equally spaced fuel compact holder openings as shown in Figure 2. Each holder opening accommodates four axially stacked fuel compacts. Thus, each capsule has three stacks by four fuel compacts per stack for a total of 12 fuel compacts per capsule. So, within the entire AGR-2 experiment capsule assembly, there are six capsules by 12 fuel compacts per capsule for a total of 72 fuel compacts. Six finite element heat transfer models were created (one for each capsule), each with a corresponding gas gap.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Wrapper for the optimization of cross-section generation in OpenMC

The "Wrapper for the optimization of cross-section generation in OpenMC" is Python-based software that runs the open-source Monte Carlo code OpenMC (https://docs.openmc.org/en/stable/ ) to generate cross-sections for any reactor geometry. The wrapper then optimizes those cross sections. The optimization aims to choose an energy group structure and a scattering representation that maximize accuracy with respect to continuous-energy results while avoiding significant computational expense. It then outputs these cross-sections in an ISOXML format readable by the Idaho National Lab code suite MOOSE (Olin William Calvin, Mark D DeHart, “Architecture for the Performance of Nuclear Fuel Depletion Calculations”, Idaho National Laboratory report, November 2019). The expected use-cases of this software include: -finding the best group structure and scattering representation for a specific reactor -testing the appropriateness of energy group structures for different reactor types -comparing energy group structures and scattering representations to each other -generating cross sections for use in deterministic codes, including ones found in the MOOSE suite The example reactor geometry included in this release is a generic reactor design, not based on any reactor in existence or in development. It was fabricated for the sole purpose of being a “testbed-geometry” upon which to develop this tool. Since the tool is designed to be generic, the nature of the test geometry is not very important, however, it is valuable to include as an example for users who are unfamiliar with developing reactor geometries for OpenMC.

Kreher, Miriam↗

DOE OSTI2015

Gamma-Ray Simulated Spectrum Deconvolution of a LaBr₃ 1-in. x 1-in. Scintillator for Nondestructive ATR Fuel Burnup On-Site Predictions

A deconvolution method for a LaBr₃ 1"x1" detector for nondestructive Advanced Test Reactor (ATR) fuel burnup applications was developed. The method consisted of obtaining the detector response function, applying a deconvolution algorithm to 1”x1” LaBr₃ simulated, data along with evaluating the effects that deconvolution have on nondestructively determining ATR fuel burnup. The simulated response function of the detector was obtained using MCNPX as well with experimental data. The Maximum-Likelihood Expectation Maximization (MLEM) deconvolution algorithm was selected to enhance one-isotope source-simulated and fuel- simulated spectra. The final evaluation of the study consisted of measuring the performance of the fuel burnup calibration curve for the convoluted and deconvoluted cases. The methodology was developed in order to help design a reliable, high resolution, rugged and robust detection system for the ATR fuel canal capable of collecting high performance data for model validation, along with a system that can calculate burnup and using experimental scintillator detector data.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2017

Application of NSUF Capabilities Towards Understanding the Emulation of High Dose Neutron Irradiations with Ion Beams

Development of new materials for current nuclear power plants (NPPs) or materials for new NPPs requires extensive research, development, and deployment programs. Central to these efforts, particularly for core internals, are studies on the microstructural and mechanical changes due to neutron irradiation. These radiation effects studies are typically completed using materials test reactors followed by expensive and time-consuming post-irradiation examinations (PIE). This sluggish approach has severely limited the nuclear materials field to develop new materials or verify materials for new reactor design use. It is also very difficult to achieve the high levels of damage proposed for future reactors. A logical approach to accelerating the field is the application of ion beams to simulate neutron damage, especially for high dose applications. Ion beams have been used for decades to study radiation effects but not to emulate high dose reactor irradiations. In this regard, new (or renewed) methods for ion beam irradiations, new developments in materials characterization, and greater understanding on the governing mechanisms controlling observed radiation effects have led to a need for further studies to refine the technique. A primary challenge is the ability to determine the ion irradiation conditions that produce microstructures and mechanical properties relevant to in-core neutron irradiation conditions. A large, multi-disciplinary, multi-laboratory program deemed the High Fidelity Ion Beam Simulation of High Dose Neutron Irradiation has been formed to assess and overcome the challenges associated with emulating high dose neutron irradiations using ion beams. Central to this program is the use of dual-beam ion irradiations and corresponding neutron irradiations within the BOR-60 fast reactor. These irradiations have produced hundreds of specimens for microstructural and mechanical characterization. A significant issue regarding the samples is the radioactivity of the neutron irradiated specimens as well as the need to characterize both irradiation types using the same equipment to maintain consistency and repeatability. The use of the Nuclear Science User Facilities (NSUF) has been central towards overcoming the mentioned issues and thereby enabling robust PIE in both the ion and neutron irradiated specimens of interest. In particular, the electron microscopes housed within the Low Activation Materials Development and Analysis (LAMDA) facility at Oak Ridge National Laboratory (ORNL) - an NSUF partner facility - has been instrumental in gathering high quality data for analysis. This paper will briefly highlight the PIE efforts using the NSUF capabilities within LAMDA.

36 MATERIALS SCIENCE↗

DOE OSTI2015

Uncertainty quantification of calculated temperatures for advanced gas reactor fuel irradiation experiments

This paper presents the quantification of uncertainty of the calculated temperature data for the Advanced Gas Reactor (AGR) fuel irradiation experiments conducted in the Advanced Test Reactor at Idaho National Laboratory in support of the Advanced Reactor Technology Research and Development program. Recognizing uncertainties inherent in physics and thermal simulations of the AGR tests, the results of the numerical simulations are used in combination with statistical analysis methods to improve qualification of measured data. The temperature simulation data for AGR tests are also used for validation of the fission product transport and fuel performance simulation models. These crucial roles of the calculated fuel temperatures in ensuring achievement of the AGR experimental program objectives require accurate determination of the model temperature uncertainties. To quantify the uncertainty of AGR calculated temperatures, this study identifies and analyzes ABAQUS model parameters of potential importance to the AGR predicted fuel temperatures. The selection of input parameters for uncertainty quantification of the AGR calculated temperatures is based on the ranking of their influences on variation of temperature predictions. Thus, selected input parameters include those with high sensitivity and those with large uncertainty. Propagation of model parameter uncertainty and sensitivity is then used to quantify the overall uncertainty of AGR calculated temperatures. Expert judgment is used as the basis to specify the uncertainty range for selected input parameters. The input uncertainties are dynamic accounting for the effect of unplanned events and changes in thermal properties of capsule components over extended exposure to high temperature and fast neutron irradiation. The sensitivity analysis performed in this work went beyond the traditional local sensitivity. Using experimental design, analysis of pairwise interactions of model parameters was performed to establish sufficiency of the first-order (linear) expansion terms in constructing the response surface. To achieve completeness, uncertainty propagation made use of pairwise noise correlations of model parameters. Furthermore, time-dependent sensitivity over the test campaign duration was obtained using an interpolation scheme over the input parameter domain. This allows computation of uncertainty for the predicted peak fuel temperatures and the predicted graphite temperatures at TC locations over the whole AGR irradiation period.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗