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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Extension and Demonstration of NEAMS Multiphysics Tools to Lead-Cooled, Sodium-Cooled, and Molten Salt Fast Reactor Applications

The SHARP toolkit is a high-fidelity reactor simulation tool developed under the U.S. Department of Energy, Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. SHARP toolkit is comprised of the neutronics module PROTEUS thermal hydraulics module Nek5000, and structural mechanics module Diablo. During FY17 and FY18, the PROTEUS and Nek5000 components of SHARP were applied to solve challenging sodium-cooled fast reactor (SFR) problems. In particular, selected hot channel factors (HCF) for a prototype metal-fueled SFR design (the AFR-100) were analyzed in high fidelity, and the “SHARP zooming capability” for SFRs was developed and demonstrated to reduce computational expense for full core problems in cases where detailed data is needed in selected fuel assemblies. After the previous success applying SHARP to challenging SFR problems, the focus in FY19 expanded to additional fast reactor applications: lead cooled fast reactors (LFR), sodium cooled fast reactors (SFR), and fast molten salt reactors (MSR). The specific technical tasks were (1) assessment of hot channel factors for LFR, for which no data currently exists, (2) implementation of a gamma transport capability in the PROTEUS solvers to extend the zooming capability to non-fueled assemblies, (3) demonstration of zooming capability in assemblies of the Versatile Test Reactor (VTR), and (4) development of a new coupled capability for simulating fast MSRs, specifically modeling precursor transport inside and outside the core. The progress on tasks (1) and (3) are discussed primarily in this document. Tasks (2) and (4) are briefly mentioned in this document, and the interested reader is referred to separate reports for detailed information. First-of-a-kind hot channel factor (HCF) estimation for LFR with high fidelity codes (PROTEUS/Nek5000) was successfully demonstrated in this study. Selected HCF were computed and compared with SFR data (AFR-100, EBR-II). The findings confirm that different reactor types, design parameters and uncertainties lead to different HCFs. Careful estimation of HCF for a specific design is necessary to obtain appropriate HCFs. In addition to improvement in HCF accuracy, high fidelity tools generate data to help the designer better understand the mechanism of the impact from these uncertainties. For example, the impact of cladding thickness manufacturing tolerance resulted in non-intuitive effects in the corner pins of the LFR assembly. This procedure of computing HCF using high fidelity models shows promise and flexibility for being repeated for any arbitrary reactor of choice, since these tools are targeted to solve many reactor types and geometries. Progress was made towards extending the previously demonstrated SHARP zooming capability to non-fueled SFR assemblies. In particular, a gamma transport capability was implemented in both high fidelity PROTEUS solvers in order to accurately account for heat deposition caused by gamma particles, which accounts for ~10% of total core power. Neutronics verification cases were carried out for a candidate Versatile Test Reactor (VTR) design using the new gamma transport capability in PROTEUS. Comparisons were made with continuous energy MCNP calculations and shown to agree well. The models for the full core design with heterogeneous control and fuel assemblies is in progress and will be completed in FY20. In the fast molten salt reactor (MSR) domain, a new capability was developed to model the flow of neutron precursors both inside and outside of the core, which impacts the effective delayed neutron fraction of the system. The PROTEUS-SN and PROTEUS-NODAL solvers were coupled to Nek5000 in order to simulate precursor flow. The workflows were demonstrated on test problems. The fast MSR modeling capability was supported under two other work packages, and comprehensive details are included in a separate report.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

A review of inherent safety characteristics of metal alloy sodium-cooled fast reactor fuel against postulated accidents

The thermal, mechanical, and neutronic performance of the metal alloy fast reactor fuel design complements the safety advantages of the liquid metal cooling and the pool-type primary system. Together, these features provide large safety margins in both normal operating modes and for a wide range of postulated accidents. In particular, they maximize the measures of safety associated with inherent reactor response to unprotected, double-fault accidents, and to minimize risk to the public and plant investment. High thermal conductivity and high gap conductance play the most significant role in safety advantages of the metallic fuel, resulting in a flatter radial temperature profile within the pin and much lower normal operation and transient temperatures in comparison to oxide fuel. Despite the big difference in melting point, both oxide and metal fuels have a relatively similar margin to melting during postulated accidents. When the metal fuel cladding fails, it typically occurs below the coolant boiling point and the damaged fuel pins remain cool-able. Metal fuel is compatible with sodium coolant, eliminating the potential of energetic fuel coolant reactions and flow blockages. All these, and the low retained heat leading to a longer grace period for operator action, are significant contributing factors to the inherently benign response of metallic fuel to postulated accidents. This paper summarizes the past analytical and experimental results obtained in past sodium-cooled fast reactor safety programs in the United States, and presents an overview of fuel safety performance as observed in laboratory and in-pile tests.

inherent safety↗
DOE OSTI2020

EVALUATION OF HOT CHANNEL FACTOR FOR SODIUM-COOLED FAST REACTORS WITH MULTI-PHYSICS TOOLKIT

The evaluation of hot channel factor (HCF) is of great significance to the quantification of safety margins for reactor designs. In this paper, HCFs for a sodium-cooled fast reactor (SFR) are evaluated with the Simulation-based High-efficiency Advanced Reactor Prototyping (SHARP) toolkit, which is developed under the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign of DOE for multi-physics reactor performance and safety simulations. The high-fidelity neutronics and thermal hydraulics solvers PROTEUS and Nek5000 in the SHARP toolkit are coupled to perform the multi-physics simulations for HCF evaluation. The HCFs induced by cladding manufacturing tolerance, fissile content mal-distribution, wire orientation and uncertainties on the cladding, coolant, and fuel properties are evaluated for a reference core SFR design (AFR-100). The HCFs calculated with the SHARP toolkit are compared to legacy HCFs for similar reactor types. The comparison demonstrates the reduction or elimination of modeling uncertainties in the calculation of HCFs using high fidelity advanced modeling and simulation tools without the need of expensive experiments. Moreover, the reduction of the uncertainties on HCFs evaluation allows an increase in nominal parameters and safety margin, which in turn improves the economic competitiveness of the SFR.

Hot channel factor (HCF)↗
DOE OSTI2016

Convective Heat Transfer in Decay Heat Exchanger of Sodium-cooled Fast Reactor

Sodium-to-sodium heat exchangers such as the decay heat exchanger (DHX) are key components of the prototype generation [V sodium-cooled fast reactor (PGSFR). The DHX transfers the decay heat generated in the reactor core to the decay heat removal loop after a reactor shutdown. Shell-and-tube type with straight tube arrangement has been widely adopted for the sodium-to-sodium heat exchanger of the sodium-cooled fast reactor. Thus liquid metal heat transfer characteristics inside the tube and around the tube bundle have to be fully understood to design the sodium-to-sodium heat exchanger and evaluate its thermal performance accurately. Plenty of studies on the convective heat transfer of the liquid metals were conducted to present the Nusselt number (Nu) correlations with respect to the Peclet number (Pe). However, those correlations still deviated from each other considerably and few of correlations for low Pe range (less than 100) have been reported. Accordingly, it is difficult for a design code developer to choose a proper heat transfer coefficient correlation for the sodium-to-sodium heat exchanger particularly in a low flow rate range. Recently, a separate effect test for the DHX has been completed using the STELLA-1 facility at KAERI according to the STELLA program which aims at establishment of separate and integral test database (DB) for the design of the PGSFR. The heat transfer performance test of the DHX was successfully conducted under major operating conditions. The objective of this work is to suggest proper convective heat transfer correlations for tube- and shell-sides of the sodium-to-sodium heat exchanger based on the test DB obtained from STELLA-1. The Pe range of the analyzed experimental data is concentrated on less then 100. The adequacy of the suggested correlations is evaluated by applying to the DHX design code, SHXSA. Based on the DHX test DB of the STELLA-1, the proper convective heat transfer correlations for the tube- and shell-side of sodium-to-sodium heat exchangers were suggested. Two correlation combinations for each side were proposed and turned out to be reasonable through the evaluation using the SHXSA design code: Suggestion no. 1: - Lyon correlation (constant = 4) for tube-side; - Lubarsky-Kaufman correlation for shell-side. Suggestion no. 2: - Modified Lyon correlation (constant = 1.45) for tube-side; - Schad-modified correlation for shell-side The Pe ranges of the DHX test DB were 53 - 225 for the tube-side and 45 - 148 for the shell-side, but most of test conditions were concentrated on the Pe range of less than 100. Two correlation suggestions can be practically used for the design of sodium-to-sodium heat exchangers. However, by utilizing more test DB, it needs to be verified which one is closer to real phenomena. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Dismantling of primary shielding with prototype machines in a compact sodium cooled reactor - 15463

The Compact Sodium Cooled Reactor Facility Karlsruhe (KNK), a prototype reactor to demonstrate the Fast Breeder Reactor Technology in Germany, was in operation from 1971 to 1991. The dismantling activities started in 1991. The project aim is the green field in 2020. Most of the reactor internals as well as the primary and secondary cooling loops have already been dismantled. The contaminated sodium inventory has already been disposed of completely. Only the highly activated reactor vessel shielding structures remain. Due to their high dose rates, these structures must be dismantled remotely. For the removal of the primary shielding of the reactor vessel, 12 stacked cast iron blocks with a total mass of 90 Mg and individual masses of up to 15.5 Mg, a remote-controlled multifunctional dismantling device (HWZ) was designed, manufactured and tested in a mock-up. Since it is not possible to remove the segments in one piece for offsite treatment, a special prototype band-saw (BWZ) had to be developed for sawing the segments both horizontally and vertically into smaller pieces. The resulting sub-segments are transferred directly into repository containers. After the successful approval of the test sequences by the authorities and intensive training of the operational staff, both prototype machines were installed inside the reactor building containment for final assembly. The dismantling started in 2013. The present paper covers the successful dismantling of the primary shielding as well as the experiences gained through the development, use and improvement of these prototype machines. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

An Analysis of Methanol and Hydrogen Production via High-Temperature Electrolysis Using the Sodium Cooled Advanced Fast Reactor

Integration of an advanced, sodium-cooled fast spectrum reactor into nuclear hybrid energy system (NHES) architectures is the focus of the present study. A techno-economic evaluation of several conceptual system designs was performed for the integration of a sodium-cooled Advanced Fast Reactor (AFR) with the electric grid in conjunction with wind-generated electricity. Cases in which excess thermal and electrical energy would be reapportioned within an integrated energy system to a chemical plant are presented. The process applications evaluated include hydrogen production via high temperature steam electrolysis and methanol production via steam methane reforming to produce carbon monoxide and hydrogen which feed a methanol synthesis reactor. Three power cycles were considered for integration with the AFR, including subcritical and supercritical Rankine cycles and a modified supercritical carbon dioxide modified Brayton cycle. The thermal efficiencies of all of the modeled power conversions units were greater than 40%. A thermal efficiency of 42% was adopted in economic studies because two of the cycles either performed at that level or could potentially do so (subcritical Rankine and S-CO2 Brayton). Each of the evaluated hybrid architectures would be technically feasible but would demonstrate a different internal rate of return (IRR) as a function of multiple parameters; all evaluated configurations showed a positive IRR. As expected, integration of an AFR with a chemical plant increases the IRR when “must-take” wind-generated electricity is added to the energy system. Additional dynamic system analyses are recommended to draw detailed conclusions on the feasibility and economic benefits associated with AFR-hybrid energy system operation.

09 BIOMASS FUELS↗
DOE OSTI2025

Consequence analyses of sabotage-induced radiological releases in sodium-cooled fast microreactors

Analysis of three sodium-cooled fast microreactors (SFMs) with thermal powers of 10, 30, and 50 MWt showed that smaller reactors result in lower radiological consequences during a postulated sabotage-induced event because of their reduced core inventory. All SFMs used U-10Zr metal fuel enriched to 15 wt% high-assay low-enriched uranium and operated until their respective effective multiplication factor (k eff ) reduced to less than 1 or until the end of their operational lifespan. Sabotage scenarios were simulated at this point, when the fuel inventory within the core contains the highest-level of radioactivity. Radionuclide core inventories were calculated using the SCALE code at shutdown and 3 days post-shutdown. Dose consequence analyses were performed for three sabotage scenarios using the RASCAL tool. As microreactor developers plan for minimal on-site or complete off-site emergency response, it remains essential to evaluate their physical protection needs and potential hazards, including assessing postulated sabotage-induced events that could become more relevant. SFM licensees should identify a credible worst-case, major accident, estimate release source terms, and perform dose consequence analyses to evaluate site-specific physical protection measures. In conclusion, this recommendation supports a risk-informed, performance-based approach, aligning with applicable regulatory requirements, i.e., 10 CFR Parts 100 and 53 rulemaking in the United States.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2024

Degradation and Failure Phenomena of Advanced Reactor Fuel Concepts: Sodium-Cooled Fast Reactor Metallic Fuel

The U.S. Nuclear Regulatory Commission (NRC) is anticipating licensing applications and commercial use of new fuel types in advanced nuclear power reactors that would be designed and built in the United States. Pacific Northwest National Laboratory (PNNL) is providing technical assistance to the NRC related to the newly proposed nuclear fuel and cladding designs that would be deployed in these reactors. This report focuses specifically on the metallic fuel that is being considered for sodium-cooled fast reactors and specifically on mechanisms that would cause damage or failure to the fuel under reactor operating conditions and design basis accident conditions. There is historic experience with both metallic and oxide fuels in experimental sodium-cooled fast reactors, but this report will focus solely on metallic fuel. Currently two U.S.-based reactor designers are engaged with the NRC in the application or pre-application review and considering a sodium-cooled fast reactor. TerraPower is engaged with NRC in pre-application review of its Natrium reactor. The current design for this fuel uses a sodium-bonded uranium-10wt% zirconium (U10Zr) fuel clad in HT9 stainless steel. ARC Clean Technology is engaged with NRC in pre-application review of its ARC-100 reactor. The current design for this fuel uses a sodium-bonded uranium-10wt% zirconium (U10Zr) fuel with steel cladding. This report will focus on this fuel system specifically, with broader information given regarding other metallic fuel systems with other stainless steel alloy claddings. To support the NRC’s readiness efforts, this report will identify and discuss degradation and failure modes of metallic fuel concepts for sodium-cooled fast reactors, including fuel performance characteristics that may not be addressed within existing regulatory documents.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Disposition Options for Sodium Cooled Fast Reactor (A White Paper)

The sodium-cooled fast reactor (SFR) design concept is one of the six classes of nuclear reactors in the GenIV initiative. SFRs are uranium or plutonium-fueled reactors operating in the fast neutron spectrum using liquid sodium as the coolant. SFRs can be designed as a breeder reactor or actinide-burning reactor in addition to operating the thorium fuel cycle. While having different fuel designs, the anticipated waste streams, and the necessary management strategies for spent nuclear fuel (SNF) and radioactive wastes from SFRs are very similar. This includes the SNF, activated sodium coolant, in-core stainless-steel components, piping, resins and filters, solidified liquid waste, contaminated equipment, and other radioactive wastes. Modern SFR designs are based on a long and rich operating history of several liquid-metal-cooled fast reactors with sodium coolant. Several of these reactors have been shut down, the fuel has been placed in safe storage, and they have undergone some degree of decommissioning. As such, there is significant experience in the management of the SNF and radioactive wastes associated with operating these reactors. This white paper will identify the definitions and regulations that apply to the safe and secure management, storage, and disposal of radioactive waste and identify the key radioactive waste streams from SFRs. Idaho National Laboratory has significant experience in the management of the SNF from the SFR predecessors. This experience should form the basis for the management and disposition efforts of the radioactive waste from any new SFR-type small modular reactor or microreactor intended for deployment at the Idaho National Laboratory Site

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Assessment of tritium effluent from Prototype Generation IV Sodium-cooled Fast Reactor

We report a tritium transport model for a pool-type sodium cooled fast reactor was developed for estimation of tritium effluents from the Prototype Generation IV Sodium-cooled Fast Reactor (PGSFR). For this purpose, the transport model developed for Experimental Breeder Reactor II and Fast Flux Test Facility was updated and validated. The tritium production rate from ternary fission and activations in PGSFR were calculated by the Argonne Reactor Computation (ARC) code suite for fast reactor analysis, and due to the potential uncertainties in source terms and variations of cold trap performance and material permeability during reactor operation, the tritium distributions and tritium effluents were estimated for several bounding cases. The upper bounding gaseous and liquid tritium release rates from PGSFR are similar to ~440.7 Ci/year and 14.7 Ci/year, respectively. The gaseous tritium effluent is about few orders of magnitude smaller than the regulatory constraint of the tritium concentration in the air. The peak tritium concentration in the steam-generator feedwater is similar to ~1.1x10 -3 mu Ci/g, which is required to be diluted prior to releasing as tritiated water. The accumulated tritium in the cold trap after five cycle operations is ~ 52,000 Ci. Because of significant conservatism in these upper bounding estimations, it is expected that the tritium effluent at the nominal case with plausible source terms would be about 2 orders of magnitude smaller than that of the upper bounding case.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Prognostic model and failure mechanisms of steam generators in Sodium-Cooled fast reactors

This paper presents a prognostic model for sodium-cooled fast reactor (SFR) steam generators (SGs). Here, the purpose of the model is to estimate the remaining useful life of SFR SGs and thus to support the decision-making of autonomous control. SFR SGs are of great interest for plant integrity due to their harsh operation environment. They operate at higher temperatures and higher coolant-to-steam pressure differences than those of current light water reactors (LWR). The severity of the SFR SG failure consequences, which include water-sodium contact, is another reason. Understanding its failure mechanisms is important for the development of its prognostic model. Based on our literature review and physics-based analysis, we concluded that creep would be a dominant degradation mode of SFR SGs due to SFR’s high-temperature and high-pressure environment. Thus, creep is the focus of the prognostic model. Various other failure modes were also investigated in this study. The mechanical fatigue due to flow-induced vibration is usually observed in early developmental phases and would not be an important issue during normal operation. The thermal fatigue due to thermal stripping is occasionally observed in other components in SFRs but does not affect SG integrity. Pure water stress corrosive crack, fretting, and so on are commonly observed in LWR SGs but are significantly less important in SFR SGs because of the high temperatures, high pressure differences, and chemical properties of liquid sodium. Based on these investigations, a prognostic model focusing on creep failures was developed. It estimates the failure probability profile by sampling the Larson-Miller parameter (LMP) and temperature and associated uncertainties through Monte Carlo methods. Two case studies were presented. The first one demonstrated the model’s capability to calculate the failure probability for a new specimen within a given time. If the tube is working under temperatures of 500C ± 3 and pressures of 15.2 MPa, which leads to LMP of 20,100 ± 50, the failure probabilities within 50, 70, and 90 years are approximately 0.1 %, 3.2 %, and 17.5 %. The second demonstrated how adjusting the workload can help to protect the integrity of the component. For an old tube reaching 99 % of its lifespan, continuing to run at temperatures of 500C ± 3 and LMP of 20,100 ± 50 leads to a failure probability of about 29.0 % within a year. If the temperature is reduced by 5C, the failure probability can be reduced to 2.5 %, and, if the pressure is also reduced such that the LMP is increased by 100, the failure probability can further be reduced to 0.12 %.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Development of a multiphysics analysis system for sodium-water reaction phenomena in steam generators of sodium-cooled fast reactors

A multiphysics analysis system for sodium-water reaction phenomena in a steam generator of sodium-cooled fast reactors was newly developed. The analysis system consists of the mechanistic numerical analysis codes, SERAPHIM, TACT, and RELAP5. The SERAPHIM code calculates the multicomponent multiphase flow and sodium-water chemical reaction caused by discharging of pressurized water vapor. Applicability of the SERAPHIM code was confirmed through the analyses of the experiment on water vapor discharging in liquid sodium. The TACT code was developed to calculate heat transfer from the reacting jet to the adjacent tube and to predict the tube failure occurrence. The numerical models integrated into the TACT code were verified through some related experiments. The RELAP5 code evaluates thermal hydraulic behavior of water inside the tube. The original heat transfer correlations were corrected for the tube rapidly heated by the reacting jet. The developed system enables evaluation of the wastage environment and the possibility of the failure propagation.

42 ENGINEERING↗
DOE OSTI2016

Shielding Design for Prototype Generation-IV Sodium-Cooled Fast Reactor

A Prototype Generation-IV Sodium-cooled Fast Reactor (PGSFR) with 150 MWe power rating is being developed jointly by the Korea Atomic Energy Research Institute (KAERI) and Argonne National Laboratory (ANL). KAERI is responsible for the overall project with the ANL contributions integrated into the KAERI activities. Shielding concept of PGSFR was developed to ensure that the safety limit on internal structural damage and working personnel received dose is not exceeded. The first part evaluated the Displacement Per Atoms (DPA) and fast neutron fluence on the structures inside the reactor vessel, such as lower grid, Upper Internal Structure (UIS), barrel, Redan, and reactor vessels. The analysis determined the thickness of shields required to satisfy the DPA and fast fluence limit. The second part of the shielding analysis evaluated the radiation dose rate above the reactor head due to the primary sodium activation inside the reactor vessel. The analysis determined the necessary thickness of reactor head and additional shield to ensure that the dose rate received by personnel working on top of the reactor is below the safety limit. The last part of the shielding analysis investigated the dose rate received by personnel working near the secondary loop. The secondary sodium is activated inside the Intermediate heat exchanger (IHX). The activated secondary sodium flows out of the reactor vessel and emits high energy photons along the way. The shielding designs both inside the reactor vessel and around the pipes of the secondary loop were investigated. The shielding concept for the PGSFR was developed to minimize radiation exposure to plant personnel and the irradiation damage of the internal structural components. The damage of the internal structural components are mainly due to the loss of the ductility by the neutron irradiation, and the activated gammas of primary and secondary sodium coolants are the major radiation sources that are required to be shielded. Thicker and denser shielding structures are favorable to reduce the radiation dose and protect personnel who work in the plant; however, which enlarges the in-vessel space to accommodate the shielding structures. Thus, in order to install the shielding structures in a reasonable reactor to satisfy the DPA or dose rate limit size, the PGSFR shielding parameters were determined from the sensitivity study, which includes the compositions of the shielding materials and dimensions. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Development of Thermoacoustic Sensors for Sodium-cooled Fast Reactors

This Final Report refers to the project “Development of Thermoacoustic Sensors for Sodium-cooled Fast Reactor Systems”, which was led by Westinghouse Electric Company (Westinghouse) and carried out in collaboration with Argonne National Laboratory (ANL) and University of Pittsburgh. Thermo-acoustic Power Sensors (TAPS) are self-powered, wireless sensors envisioned for measuring key parameters, such as local temperature and neutron flux, in a nuclear reactor core. This project was intended to specifically investigate their applicability to Sodium-cooled Fast Reactors (SFR). TAPS are non-invasive (wireless) and passive (self-powered) devices. The passivity derives from their ability to use conditions that “naturally” exist in a nuclear reactor, such as gamma and neutron flux, as power sources. They generate oscillating pressure waves (i.e., sound waves) which, with a frequency and amplitude dependent upon these conditions, can travel through the core and associated structures, and reach the outside of the reactor vessel where a properly designed network of receivers can detect and interpret them. These receivers require a very small amount of power which, during loss of power events, can be provided for example by harvesting gamma radiation energy, thus resulting in a monitoring system that can function both during normal operation and during loss of power events. The project aimed at TAPS development through a series of tasks which are listed and briefly discussed as follows. TASK 1 – Sensor hardware design Subtask 1a: Assessment of sensor applications to SFRs Subtask 1b: Development of sensor functional requirements Subtask 1c: Definition of sensor hardware design specifications Task description: TAPS design was informed by considerations on their application (Subtask 1a), both the ultimate one in an SFR and the actual one in the ANL testing facilities that was intended to be used in support of the project. Considerations were made to identify optimum sensor design features that optimize the sensor size, materials, and output signal, for installation inside an SFR core. These considerations led to the development of Functional Requirements (Subtask 1b) and Design Requirements (Subtask 1c). TASK 2 – Sensor Hardware Manufacture Subtask 2a: Sensor hardware construction drawing development Subtask 2b: Sensor manufacture and assembly Task description: TAPS technical drawings were developed (Subtask 2a) using the Design Requirements established under Task 1. Subsequently, in spite of some problems which ultimately caused the program to be delayed, TAPS manufacturing was completed based on drawings (Subtask 2b). TASK 3 – Development of TAPS Signal Measurement System and TAPS Testing in Water Subtask 3a: Design, assembly and testing of signal measurement system, and TAPS testing in water Subtask 3b: Signal prediction-correction methodology development Task description: An assessment was performed on the techniques that can potentially be used to detect the signals emitted by the TAPS, e.g. a fiber-optic based acoustic signal measurement system, a laser vibrometer system, or an accelerometer-based system. The most suited technology, i.e. the accelerometer-based system, was developed further, and tested in water (Subtask 3a). Moreover, efforts were made to develop the methodology required to determine the actual system temperature and neutron flux distribution using differences between the measured and predicted TAPS responses (Subtask 3b). TASK 4 – Sensor System Testing in Sodium Subtask 4a: Test plan development Subtask 4b: Design, assembly and testing in small-scale sodium facility Subtask 4c: Design, assembly and testing in large-scale sodium and structures facility Task description: Upon proper test plan development (Subtask 4a), the fabricated TAPS was planned to be tested in sodium, by using two sodium facilities at ANL having different size and different purpose. The Under Sodium Viewing (USV) small-scale facility was intended to be used to investigate the effect of sodium on the sensor and its performance (Subtask 4b). The Mechanism Engineering Test Loop (METL) large-scale facility was instead intended to be used to assess the additional effect of prototypical SFR structures, such as fuel assembly mockup or parts of the core restrain structure, on sensor performance (Subtask 4c). As discussed in Section 3.2.2.7, unexpected issues during the TAPS manufacturing process resulted in some activities being delayed, with the TAPS and USV facility developed to the point to be ready for testing in sodium, however without the possibility to actually perform such testing (including the testing in METL) due to the end of the program’s performance period. Overall, through the development and testing (in water only) of two TAPS devices (a First-Generation TAPS followed by an optimized Second-Generation TAPS), the project confirmed the capability of this technology to generate acoustic signals proportional to temperature, which can be detected through a network of accelerometers identified as the best-suited type of receivers for acoustic signal detection. Moreover, the project also developed a computational model to predict the characteristics of the acoustic signals being generated, which combines thermal analysis of the TAPS with Finite Element Modeling (FEM)-aided acoustic characterization of the system. This model was benchmarked against experimental data collected during the project and, although general agreement was obtained, some limitations of the modeling methods were identified, which require additional development. Additional testing is needed in order to assess the effect, on TAPS operation and performance, of environmental changes resulting from the transition from water to liquid sodium. Such testing, which is suggested to be performed in the future, should look specifically at 1) both the effect resulting from the different thermoacoustic behavior of sodium (relative to water) and the effects of higher temperature on TAPS performance, and 2) the performance of the sensor-receiver system when multiple TAPS are used simultaneously and prototypical reactor structures are positioned in the testing environment. The latter testing is needed to assess the effects that potential signal attenuation/ distortion phenomena, as well as potential interference between signals emitted simultaneously, have on the performance of the technology for ultimate application in a nuclear reactor.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Alternative Fabrication of Volatile Metal Fuel Slugs for Sodium-cooled Fast Reactor

In order to develop alternative fabrication method of fuel slugs for preventing the evaporation of volatile elements such as AM,U-10wt.%Zr-Mn fuel slugs for SFR (: Sodium-cooled Fast Reactor) have been fabricated by modified casting method and characterized to evaluate the feasibility of the alternative fabrication. Metal fuels such as U-Zr/U-Pu-Zr alloys have been considered as a nuclear fuel for a sodium-cooled fast reactor (SFR) related to the closed fuel cycle for managing minor actinides and reducing a high radioactivity levels since the 1980's In order to develop innovative fabrication method of metal fuel slugs for preventing the evaporation of volatile elements such as Am, modified casting under inert atmosphere has been applied for metal fuel slugs for SFR. Alternative fabrication method of fuel slugs has been introduced to develop an improved fabrication process of metal fuel for preventing the evaporation of volatile elements. In this study, U- 10wt.%Zr-Mn fuel slugs for SFR have been fabricated by modified casting method and characterized to evaluate the feasibility of the alternative fabrication method. Alternative casting such as modified casting has been applied to develop fabrication method of fuel slugs for preventing the evaporation of volatile elements such as Am. U-10wt.%Zr-Mn containing a volatile surrogate Mn fuel slugs was soundly fabricated under inert atmosphere with dimensions of Φ5 - L 250 mm. Mass fraction of fuel loss was so low, up to 0.2%. Mn element was most recovered with prevention in evaporation of Mn. It was seen that the losses of volatile Am can be effectively controlled to below detectable levels using modest pressure. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Overall Reactor System Description of Prototype Gen-IV Sodium Cooled Fast Reactor

A fast reactor system has been recognized as one of the most promising options for transmutation of the radioactive wastes discharged from spent nuclear fuel of light water reactors (LWRs). That is owing to the intrinsic features of a fast reactor, i.e., high neutron energy and abundant excessive neutrons in the core. In this context, the Korea Atomic Energy Commission (KAEC) authorized the R and D action plan for the advanced sodium cooled fast reactor (SFR) connected with the pyro-process to provide a consistent direction to long-term R and D activities in December 2008. The long term plan was updated in November 2011 in order to refine the plan and to consider the available budget for SFR. The revised milestones include the completion of a specific design of a prototype Gen IV sodium cooled fast reactor (PGSFR) by 2017, specific design approval by 2020, and construction by 2028. The R and D activities on SFR had been concentrated on the basic and strategic key technology developments before 2012 and went to engineering project for licensing and construction of PGSFR from 2012. The development of PGSFR is now in the first design phase and will go toward to the second design phase in 2016. All the high level design concepts of systems, structures and components (SSCs) are determined and incorporated into the preliminary safety information document (PSID), which includes basic design requirements and descriptions and the results of safety analysis for the representative accident scenarios. The PSID will be a base material for a pre-review of the PGSFR safety. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

CFD Evaluation of Pressure Change Along Coolant Passages in Sodium-Cooled Fast Reactor with Nek5000

To support the design efforts of advanced sodium-cooled fast reactors (SFRs), a series of computational fluid dynamics (CFD) simulations are performed to investigate the pressure change along various flow passages in the proposed SFR system. The simulations are carried out with the state-of-the-art spectral element flow solver, Nek5000. Two specific case studies are presented in this paper: the flow exiting the axial neutron reflector channels and the flow entering the fuel pin bundle. Due to the high Reynolds numbers expected, a Reynolds-averaged Navier-Stokes (RANS) approach is necessary to model the turbulence. A newly developed regularized k-omega RANS model is adopted in the related CFD calculations. The first case study explores the effect of Reynolds number on the pressure change when flow exits the reflector channels. The pressure change in this case has two major contributors: the change due to wall friction and the Bernoulli effect. It is noted that the nondimensional pressure loss follows a log-linear trend up to Re = 10 5 , and then the trend is flattened. In the second case study, the advanced NekNek coupling capability is tested where an integral domain can be divided into multiple subdomains with coupling interfaces, which would greatly ease the meshing process of complex engineering geometries and potentially save computational resources. The preliminary results obtained so far confirm the consistency between the NekNek results and those produced by regular Nek5000 simulation. The presented work demonstrates the readiness and flexibility of the related CFD techniques, which is part of the broader effort to leverage cutting-edge CFD to inform the advanced nuclear reactor designs.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Upsampling Monte Carlo Reactor Simulation Tallies in Depleted Sodium-Cooled Fast Reactor Assemblies Using a Convolutional Neural Network

The computational demand of neutron Monte Carlo transport simulations can increase rapidly with the spatial and energy resolution of tallied physical quantities. Convolutional neural networks have been used to increase the resolution of Monte Carlo simulations of light water reactor assemblies while preserving accuracy with negligible additional computational cost. Here, we show that a convolutional neural network can also be used to upsample tally results from Monte Carlo simulations of sodium-cooled fast reactor assemblies, thereby extending the applicability beyond thermal systems. The convolutional neural network model is trained using neutron flux tallies from 300 procedurally generated nuclear reactor assemblies simulated using OpenMC. Validation and test datasets included 16 simulations of procedurally generated assemblies, and a realistic simulation of a European sodium-cooled fast reactor assembly was included in the test dataset. We show the residuals between the high-resolution flux tallies predicted by the neural network and high-resolution Monte Carlo tallies on relative and absolute bases. The network can upsample tallies from simulations of fast reactor assemblies with diverse and heterogeneous materials and geometries by a factor of two in each spatial and energy dimension. The network’s predictions are within the statistical uncertainty of the Monte Carlo tallies in almost all cases. This includes test assemblies for which burnup values and geometric parameters were well outside the ranges of those in assemblies used to train the network.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗