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Meta-Analysis of Advanced Nuclear Reactor Cost Estimations

Supporting Data can be downloaded at: https://gain.inl.gov/content/uploads/4/2024/06/INL-RPT-24-77048-R1.xlsx Nuclear energy is a critical cornerstone of the current United States clean energy supply and may play a larger role in the future in support of a transition to a net-zero economy. The current fleet of nuclear reactors predominantly consists of large light-water reactors (LWRs), while many of the reactor designs under consideration are smaller and/or different technologies. Because these new designs have not yet been built, there is a high degree of uncertainty associated with their cost. This complicates energy-planning efforts because cost projections are not always standardized, consistent, and centralized in an easily accessible location. To help support energy planning in the US, this report provides advanced nuclear cost ranges using a transparent methodology along with other relevant information that can be used to help support decision making and energy planning. The purpose of this work was to conduct a methodical process for cost evaluation using only public information that was vetted with the end-goal to provide reference cost projections for nuclear energy. To provide a solid basis for these values, the approach and assumptions are explicitly laid out throughout the report allowing any user of the data to challenge or reconsider them. Because future US nuclear-reactor costs are still unknown due to little recent observed data, the report opted to compile a comprehensive list of bottom-up estimates and evaluate averages/trends within the data to identify reference ranges. This was deemed preferable to opining on the robustness or validity of one cost estimation versus another. To that end, the work evaluated thousands of lines of cost subaccounts from several bottom-up cost estimates. A wide variety of different reactor types captured in the data are of various sizes and technologies. Some of these reactors will be representative of advanced reactors under development while others will not. Thus, the results here are dependent on the data that are available and the accuracy of the estimates that are used. Each bottom-up estimate was reviewed to determine whether it was complete. Incomplete data sets were corrected to ensure an adequate basis of cross-comparison. The report is not without limitations and should be interpreted as an initial step to develop cost ranges for nuclear technology. Ultimately, future work can build upon the methodology with refined cost estimates to reduce uncertainty. US-based overnight capital cost (OCC) estimates were compiled from extensive data sets into ranges for both large and small reactor sizes for 2030. To project the cost declines over time, learning rates were sampled from literature sources. No SMRs were previously built; hence, learning rates based on bottom-up approaches (e.g., by quantifying the impact stemming from fabrication of different components, modular work, site construction, commissioning) were prioritized. For larger reactors, actual learning rates from deployments were used to project future costs (adjusted to account for standardization or lack thereof between designs). Other costs included are fixed and variable operations and maintenance costs. The final variables were capacity factors and ramp rates to support energy planning.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Advanced Test Reactor Long-Term Asset Management Accomplishments 2016 to 2022

Advanced Test Reactor Long-Term Asset Management Accomplishments report contains photographs and information on completed projects or completed phase of a project for 2022. This book will be presented for the year-end report in Washington, D.C. as well as delivered to our clients at the Navy Nuclear Laboratory.

99 GENERAL AND MISCELLANEOUS↗

MEASUREMENT OF THE NEUTRON FLUENCE-RATES IN THE ADVANCED TEST REACTOR

The axial neutron fluence-rate distribution at the Advanced Test Reactor (ATR) is determined using a combination of pure nickel and cobalt-aluminium alloy dosimeter wires. Most irradiation programs at ATR need to know the fast neutron fluence-rate surrounding their experiment to correlate post-irradiation examination results with neutron dose. The activated dosimeter wires, which extend beyond the full fuelled region of the reactor core, are highly radioactive following the full cycle in ATR and must therefore be measured in the ATR canal. The system that is employed for these measurements is mechanical scanning mechanism is used to scan the activated wires across the face of a columnated shield and a NaI detector. This detector is not sufficiently calibrated to provide absolute activity measurements, therefore, the count-rate is recorded and normalized to the highest activity for that wire. Then, an approximately 3-mm long portion of the wire is removed from the highest activity region in the wire. The wires are then transferred, and an absolute activity measurement is made using a calibrated ionization chamber. These measurements are combined with corrections for decay, neutron energy spectrum, burnup, dosimeter composition, reactor power fluctuations, and reactor outages to determine the average axial neutron fluence-rate at various locations within ATR during the cycle.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Advanced Test Reactor Safety Basis Update for Gas-Cooled Experiments

The Advanced Test Reactor (ATR) supports neutron irradiation of several types of experiments. One such experiment type is referred to as a gas leadout. Gas leadout experiments actively flow gas through the experiment which allows for active temperature control. It also allows for in-situ data of the experiment. For example, fission gas migration through a fuel sample can be monitored via activity of the sweep gas. Historically, ex-pile equipment and fission product monitors were housed in shielded ATR cubicles. Due to other facility updates, cubicle space is no longer available for gas leadout experiment equipment. To support continued operation of gas leadout experiments, ATR completed a safety basis update that supports a new housing for leadout equipment that may process potentially contaminated gas. In addition to the structure and associated equipment, technical safety requirements regarding handling and storage of experiments needed to be revised to support fueled gas leadout experiments and associated outage configurations. The safety basis update addressed the full lifecycle of these experiments, including experiment movement and interim storage, and credible abnormal events such as failures or leaks in contaminated gas tubing in occupied areas. This paper discusses the completed analyses performed to support the safety basis update associated with gas leadout experiments, including thermal-hydraulic evaluation, probabilistic analysis, and dose consequence analyses.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

In-Canal Assay of High Specific Activity 60 Co at the Advanced Test Reactor

Production of high specific activity (HSA) 60 Co has recently resumed at the Advanced Test Reactor (ATR) at the Idaho National Laboratory. The technical steps of performing in-canal assay of HSA 60 Co targets at the ATR are described herein. The HSA targets are assayed on a regular basis, between cycles at the ATR, in order to assess the progress of activation. The targets are also assayed at the conclusion of the irradiation in order to provide activity estimates for the distributor and to be used for safety and shipping evaluations. These target assay activities must take place in the ATR canal to provide sufficient radiological shielding. Furthermore, a specialized assay fixture is used in conjunction with custom 60 Co radiation standards to assess the irradiated target. The specific activity of each irradiated target was determined by first measuring the 60 Co standards to determine a sensitivity factor. Then, each irradiated target was measured, and the integrated measurement was multiplied by the sensitivity factor to determine the total activity of the irradiated target. Finally, a correction factor was determined to adjust the reported activity, accounting for differences in the physical geometries of the standards and targets. These methods were used to accurately assay the gross activity of nine irradiated 60 Co targets, two of which were delivered to the distributor where confirmatory hot-cell assays were performed verifying the accuracy of the in-canal assay method.

07 ISOTOPE AND RADIATION SOURCES↗

Initial Observations from AGR 5/6/7 Capsule 1

The fourth and final irradiation experiment in the Advanced Reactor Technologies (ART) Advanced Gas Reactor (AGR) fuel development and qualification program is designated as AGR-5/6/7. Data collected from the fabrication, irradiation, and post-irradiation examination (PIE) of this tristructural isotropic (TRISO) fuel are intended to serve as the primary data set for the qualification of this fuel for use in high-temperature gas-cooled reactors (INL 2021, Collin 2018b). However, data collected from the three preceding irradiations (i.e., AGR-1, AGR-2, and AGR-3/4) may also be used to supplement data collected from AGR-5/6/7. All components of the AGR-5/6/7 fuel (i.e., UCO kernels, TRISO coatings, and fuel compacts) were produced on an engineering scale at BWXT (Lynchburg, Virginia USA) according to the fuel specification (Marshall 2017). This fuel was irradiated in the northeast flux trap (NEFT) at the Advanced Test Reactor (ATR) at Idaho national Laboratory (INL) from February 16, 2018 to July 22, 2020 (Pham et al. 2021). Measurements in the fission product monitoring system (FPMS) indicated unexpected and significant numbers of failures of TRISO particles in Capsule 1 near the end of the sixth irradiation cycle (ATR Cycle 166A). In the fourth cycle (ATR Cycle 164B) and beyond, the sweep gas flow became very low (presumably from degradation of the capsule gas outlet line via an unidentified mechanism), and the program deliberately isolated Capsule 1 from gas flow periodically. In later cycles, attempts to reestablish any kind of flow in Capsule 1 were unsuccessful. With little or no flow through Capsule 1, FPMS measurements and enumerations of failed particles in Capsule 1 were difficult or impossible as was the ability to control the helium/neon gas mixture used for temperature control. Gas flows and fission gas activity in the effluent gas from the other AGR-5/6/7 capsules were also impacted by the Capsule 1 gas flow issues and the large increase in fission gas released from the Capsule 1.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Cross-Code Verification of Neutronics Analysis Tools at INL Applied for 238 Pu Production in the Advanced Test Reactor

Here, analyses are completed for experiments prior to experiment irradiation in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). Various codes are used to qualify all experiments planned for insertion in the reactor, thereby ensuring that all safety and programmatic requirements are satisfied preirradiation. Among the common experiment analysis tools at INL are MCNP5 coupled to ORIGEN2 (MOPY) and MC21. MOPY uses MCNP5 for transport calculations along with calculations for fluxes and select reaction rates, and then ORIGEN2 handles the step-by-step and postirradiation depletion. MC21 handles all in-reactor transport and step-by-step, during-irradiation, depletion calculations, and then ORIGEN (SCALE 6.2.3) is used for decay and dose calculations postirradiation. The MOPY results, along with those obtained via two variations of the MC21 model, were compared in terms of 238 Pu production in the ATR’s H10 position. For the MOPY model, the MC21 model utilizing the HELIOS-based fission product (FP) library, and the MC21 model utilizing the expanded 1300 FP library, the during-cycle irradiation in-core heating results were sufficiently equivalent; however, the MOPY model and the MC21 model with the HELIOS library showed some differences relating to the respective FP libraries. Ultimately, the MC21 model with a 1300 FP library produced the most consistent results throughout the cycle, whereas the MC21 model that utilized the (smaller) HELIOS library was able to handle during-irradiation analysis but lacked certain short-lived FPs that significantly contributed to the total decay heat at shutdown. MOPY, on the other hand, was found to overpredict fission gas production, as a result of limitations in the ORIGEN2 code.

ATR↗

Application of Fuel Depletion Chain Simplification to Experiment Analysis in the Advanced Test Reactor

An irradiation experiment analysis can be informed by high-fidelity reactor engineering depletion results, but this comes at a computational cost. Applying depletion chain simplification to the advanced test reactor driver fuel before performing experiment depletions permits their programmatic parameters to be calculated faster, with a small penalty to accuracy. Here, this work contrasts the results of two irradiation experiments with different neutronic characteristics. Overall, the simplified nuclide library produced using a simple one-group microscopic cross-section library for a pressurized water reactor in the depletion chain simplification process performed comparably in terms of accuracy and runtime to the simplified nuclide library produced using a three-group microscopic cross-section library generated specifically for the advanced test reactor experiments being modeled. This is attributed to the additional nuclides and transmutation pathways preserved in the one-group cross-section library, which has data for 297 nuclides, compared to the three-group cross-section library, which has data for 217 nuclides. This indicates that a cross-section library with more nuclides is better than a cross-section library with fewer nuclides for the depletion chain simplification process, even if the cross-section library with fewer nuclides better represents the flux spectrum of the system being considered.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Neutrino-detector design for safeguarding small modular reactors

Nuclear reactors have long been a favored source for antineutrino measurements for estimates of power and burnup. With appropriate detector parameters and background rejection, an estimate of the reactor power can be derived from the measured antineutrino event rate. Antineutrino detectors are potentially attractive as a safeguards technology that can monitor reactor operations and thermal power from a distance. Advanced reactors have diverse features that may present challenges for current safeguards methods. By comparison, neutrino detectors offer complementary features, including a remote, continuous, unattended, and near-real-time monitoring capability, that may make them useful for safeguarding certain classes of advanced reactors. This study investigates the minimum depth and size of an antineutrino detector for a small modular reactor to meet safeguards needs for advanced reactors. Furthermore, extrapolating performance from several prior reactor antineutrino experiments, this study uses an analytical approach to develop a possible design for a remote antineutrino-based monitoring device.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Report on Year-3 of Water NSTF Matrix Testing: Two-Phase Parametric Test Series

The Natural convection Shutdown heat removal Test Facility (NSTF) is large scale test facility constructed at Argonne National Laboratory to generate NQA-1 qualified validation data for passive decay heat removal systems in advanced reactor concepts. With support from the Department of Energy (DOE) offices of Advanced Reactor Technologies (ART) and Advanced Reactors Regulatory Development (RD), this facility reflects key features of a ½ scale, water-based, Reactor Cavity Cooling System (RCCS) and is intended to study the behavior, bound performance, and ultimately guide design decisions for passive decay heat removal systems. Since the facility also retains all aspects common to a fundamental boiling water thermosiphon, the generated experimental data sets are well poised to advance the understanding of fundamental natural circulation boiling and two-phase flow phenomena.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

At-power subcritical multiplication in the advanced test reactor

The degree to which a given physical subdivision of a critical reactor acts as a neutron source for other lobes is not analytically characterized for general application. ATR and ATRC both operate at power, and individual power-producing lobes rely on each other as neutron sources in order to maintain constant power, which in general requires either exactly critical multiplication within a reactor or an external neutron source. Here, this work shows that lobe powers in ATR or ATRC can be related with subcritical multiplication theory, demonstrating that the lobes are not individually critical. This explanation is significant for ATR due to the desire to irradiate a large variety of experiments simultaneously, each having its impact on the core neutron population. This work also establishes the lack of relationship between experiment reactivity worth and cross-sections calculated from fundamental nuclear data. For any physical subdivision of any other critical reactor, it is likewise true that the subdivision undergoes only subcritical multiplication.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Measuring changes in environmental radiological background from construction of an advanced nuclear reactor testing facility

We present a method to survey and track changes in the environmental radiological background during the construction and operation of advanced nuclear reactor facilities. We discuss the results of two surveys of the environmental gamma-radiation background at NEXT Lab, an advanced nuclear reactor research facility on the campus of Abilene Christian University, prior to the introduction of radioactive material. In both surveys, the observed radiation dosage rates are low, with the highest rates at 15% of the average total radiation dosage rate for the United States. We observe ≈20% changes in the radiological background of the property in locations where the environment was changed by construction and ≈20% variations within the facilities correlated with variations in building materials.

Environment↗

Use Cases and Model Development of Thermal Storage Coupling for Advanced Nuclear Reactors

This report discusses the different options for coupling thermal energy storage (TES) systems to advanced nuclear power plants (A-NPPs) in order to enable flexible and hybrid plant operation. An advanced light-water reactor (ALWR) and a high-temperature gas-cooled reactor (HTGR) were selected as the initial use cases for demonstrating a thermally balanced energy storage coupling design for thermal power extraction. Cost functions for the A-LWR were derived from the fully balanced models that were developed based on three different coupling options with three different thermal energy bypass ratios. For the next steps, cost functions for the HTGR will also be derived, and additional nuclear reactors (e.g., a liquid-cooled fast reactor [LFR] or molten-salt reactor [MSR]) will be evaluated for coupling with TES in similar fashion, including the evaluation of their steady-state condition models and cost functions. The models presented herein showcase several design considerations, focusing on optimal deployment methodologies for achieving steady-state operation with minimum disruption to the nuclear power generation cycle. This report presents the results of steady state models developed using Aspen HYSYS®, wherein the thermal energy bypass for an NPP-TES coupling was varied up to 50%. The various components were sized using Aspen Process Economic Analyzer (APEA) and Aspen Exchanger Design and Rating (EDR), when applicable. Cost functions from these models were developed using the latest publicly available data obtained from APEA V11. The current steady-state models and cost functions provide a baseline for additional work focusing on dynamic operation and process optimization by using Idaho National Laboratory (INL)’s Framework for Optimization of Resources and Economics (FORCE) tools to evaluate the technoeconomic viability and transient operations of TES-coupled A-NPPs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Interoffice Memorandum RE-04-21 "Advanced Test Reactor Power History Through Cycle 169A-1"

Table 1 lists the Advanced Test Reactor (ATR) N-16 constrained power history data since the Beryllium VI Core Internals Changeout (CIC) Cycle 134A-1 through Cycle 169A-1. The powers tabulated for Cycles 159A-1, 163A-1, and 167A-1 are higher than powers observed during operation, because powers are computed from cycle exposure, which for these cycles includes operation during the low-power “soak” portion of each cycle; whereas cycle length in Effective Full Power Days (EFPD) includes only the high-power “casualty” portion occurring after NF is established. Table 2 lists the accumulated N-16 lobe and total core exposure, as obtained from the ATR Data Acquisition System (RDAS) for Cycles 134A-1 through 169A-1. Table 3 lists the startup and shutdown dates and times, as obtained from logbooks or RDAS, depending on availability. The ATR power history prior to Cycle 134A 1 is presented in references (a) through (d).

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Verification of the 3-Region Advanced Test Reactor MCNP Model

The verification of the 3-region homogenized fuel Advanced Test Reactor MCNP model. The 3-region model was compared to the 19-plate model found in the 94-CIC report. Flux tallies, energy deposition tallies, and quarter core mesh tallies were used to compare the two models. The 3-region model needed updating in order to make good comparisons between the models. The percent error from the flux and energy deposition tallies data shows that experiment positions inside the flux trap have higher errors than positions outside the fuel ring. The standard deviation data obtained from the mesh tallies shows that the two models agree within two standard deviations throughout the reactor. It is concluded that the model works adequately for what it is used for.

99 GENERAL AND MISCELLANEOUS↗

Verification of the 3-Region Advanced Test Reactor MCNP Model

The verification of the 3-region homogenized fuel Advanced Test Reactor MCNP model. The 3-region model was compared to the 19-plate model found in the 94-CIC report. Flux tallies, energy deposition tallies, and quarter core mesh tallies were used to compare the two models. The 3-region model needed updating in order to make good comparisons between the models. The percent error from the flux and energy deposition tallies data shows that experiment positions inside the flux trap have higher errors than positions outside the fuel ring. The standard deviation data obtained from the mesh tallies shows that the two models agree within two standard deviations throughout the reactor. It is concluded that the model works adequately for what it is used for.

99 GENERAL AND MISCELLANEOUS↗

SULI Internship Deliverables (Flow Orifice Design: Measuring Degas Flow of Experiment Loops at the Advanced Test Reactor)

These deliverables present a design for a new orifice flow meter to be installed in three experiment loops at the Advanced Test Reactor. The current configuration of the experiment loop system has an orifice that is not properly sized to the low flow rates of the degas line of the pressurizer system, causing inaccurate flow measurements. Under consideration of flow conditions and system constraints, an iterative procedure was utilized to determine the optimal sizing for a new orifice. This orifice is designed to accurately and precisely measure the flow rate of the degas line, which will consequently improve control of the pressure and chemistry of the loop pressurizer system. The scope of this project includes recommendations for orifice sizing, material, flange type, face type, and installation location on the loop piping spools. The design is suitable for experiment loops 1D-N, 2B-SE, and 2D-SW and meets all necessary standards.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗