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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 127 records · Page 7
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

On the Elevated Temperature Thermal Stability of Nanoscale Mn-Ni-Si Precipitates Formed at Lower Temperature in Highly Irradiated Reactor Pressure Vessel Steels

Atom probe tomography (APT) and scanning transmission electron microscopy (STEM) techniques were used to probe the long-time thermal stability of nm-scale Mn-Ni-Si precipitates (MNSPs) formed in intermediate and high Ni reactor pressure vessel steels under high fluence neutron irradiation at ≈320 °C. Post irradiation annealing (PIA) at 425 °C for up to 57 weeks was used to determine if the MNSPs are: (a) non-equilibrium solute clusters formed and sustained by radiation induced segregation (RIS); or, (b) equilibrium G or Γ 2 phases, that precipitate at accelerated rates due to radiation enhanced diffusion (RED). Note the latter is consistent with both thermodynamic models and x-ray diffraction (XRD) measurements. Both the experimental and an independently calibrated cluster dynamics (CD) model results show that the stability of the MNSPs is very sensitive to the alloy Ni and, to a lesser extent, Mn content. Thus, a small fraction of the largest MNSPs in the high Ni steel persist, and begin to coarsen at long times. These results suggest that the MNSPs remain a stable phase, even at 105 °C higher than they formed at, thus are most certainly equilibrium phases at much lower service relevant temperatures of ≈290 °C.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Numerical Simulation of a Natural Convection–Driven Air-Cooled Reactor Cavity Cooling System Experiment

Ensuring the efficient removal of decay heat from the reactor vessel is essential for the safety of advanced reactor technologies. Several Generation-IV concepts incorporate variations in the reactor vessel cooling systems to achieve this objective. High-temperature gas-cooled reactors utilize a reactor cavity cooling system (RCCS), a passive ex-vessel system designed to operate without active components or external power during accident conditions. The RCCS removes decay heat primarily through radiative and convective heat transfer mechanisms. Here, this study presents a comprehensive validation of a computational fluid dynamics Reynolds-averaged Navier-Stokes model for the University of Wisconsin-Madison air-cooled RCCS facility. Validation was conducted for both high- and low-power natural convection cases under a uniform heating profile. Near-wall resolution was found to be critical for accurately modeling natural convection in the RCCS; employing an all-𝑦 + wall treatment resulted in wall temperature discrepancies exceeding 50 °⁢𝐶 compared to a wall-resolved mesh. Thermal-hydraulic behaviors under natural and forced convection conditions were compared within the heated cavity and RCCS. A turbulence model sensitivity analysis indicated that low-Reynolds number k-ɛ, k-ω shear stress transport (SST), and Reynolds stress transport models produce similar wall temperature predictions. A buoyancy modeling sensitivity study revealed that the Boussinesq approximation significantly underpredicted thermal-hydraulic behavior in the RCCS. Based on these findings, modeling recommendations are provided. The validated data set along with identified sensitivities refine the modeling of natural convection in the RCCS. The information produced by this study supports RCCS design, optimization, and safety evaluations, enabling the calibration and verification of reduced-order thermal-hydraulic models.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Light Water Reactor Sustainability Program: Report on the Completion of the Machining of Zion Unit 1 Reactor Pressure Vessel Blocks into Mechanical and Microstructural Test Specimens and Chemical Analysis Coupons

The decommissioning of the Zion Units 1 and 2 Nuclear Generating Station in Zion, Illinois, presented a unique opportunity for developing a better understanding of materials degradation and other issues associated with extending the lifetime of existing nuclear power plants (NPPs) beyond 60 years of service. In support of extended service and current operations of the US nuclear reactor fleet, the Oak Ridge National Laboratory (ORNL), through the Department of Energy (DOE), Light Water Reactor Sustainability (LWRS) Program, coordinated with Zion Solutions, LLC, a subsidiary of Energy Solutions (ES), the selective procurement of materials, structures, and components, from the decommissioned reactors including multiple segments of the Zion Unit 1 Reactor Pressure Vessel (RPV).

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Neutron-Gamma flux and Dose calculations in a Pressurized Water Reactor (PWR)

The Institute of Radiological Protection and Nuclear Safety (IRSN) acts as technical support to the French nuclear safety authorities. In this context, one of its missions is to improve nuclear safety through the enhancement of knowledge regarding operating nuclear power reactors. The issues concerning the aging and the characterisation of the reactor vessel require knowledge of the neutron and gamma radiation environment on the vessel and beyond. Since few years, ENEA (Energia Nucleare e Energie Alternative, i.e. Nuclear Energy and Alternative Energy agency) and IRSN collaborate on Monte-Carlo simulations and Variance Reduction Techniques (VRT), which are used to determine neutron and gamma spectra in and out of the core. Additionally, IRSN participates in the DISCOMS project, which stands for 'Distributed Sensing for Corium Monitoring and Safety', aiming at examining the status of the third barrier of confinement, in order to define possible mitigation strategies in case of a severe accident, namely: i) reactor pressure vessel breakthrough and corium relocation outside the vessel, ii) reactor pit concrete erosion and iii) corium cooling. One of the IRSN tasks in the frame of the DISCOMS project is to provide the radiation level in terms of neutron and gamma doses and flux energy spectra at various locations. This paper summarizes the study performed on the radiation transport in the Tihange-I Belgian nuclear reactor and contributes to several issues requiring the knowledge of the radiation environment in and outside the reactor vessel. This paper gives an overview of the work presented at the ICRS-13 and RPSD-2016 conference (for more details see the proceedings paper)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

SOME USEFUL MECHANICAL PROPERTY CORRELATIONS FOR NUCLEAR REACTOR PRESSURE VESSEL STEELS

The use of correlations is common in the research and development arena of the nuclear industry with the realization that some applications with direct implications to safety demand a more rigorous approach. Most correlations involve the relationship between two experimental properties, such as that between hardness and tensile strength. There are others that are much more complicated and are often designated models because they incorporate physically-based knowledge; examples of this are predictive correlations for irradiation-induced embrittlement of reactor pressure vessels (RPV). The objective of this paper is to collect and discuss many of the commonly used correlations for applications to nuclear RPVs. This paper identifies and discusses various correlations that relate easily measured properties to properties that are more difficult, more time consuming, or more expensive to measure. In the case of irradiated RPV materials, irradiation-induced changes in easily measured properties are related to the changes in those more difficult to measure. It is noted that recognition and understanding of the uncertainties associated with all correlations is highly important.

Nanstad, Randy K.↗
DOE OSTI2015

Analysis of N-16 concentration in primary cooling system of AP1000 power reactor

Nitrogen-16 (N-16) is one of the radiation safety parameter on the primary reactor system. The activation product, N-16, is the predominant contributor to the activity in the reactor coolant system during reactor operation. N-16 is activation product derived from activation of O-16 with fast neutron based on {sup 16}O(n,p){sup 16}N reaction. Thus study is needed and it performs to determine N-16 concentration in reactor coolant (primary coolant) in supporting radiation safety. One of the way is using analytical methode based on activation and redecay princip to obtain N-16 concentration. The analysis was performed on the configuration basis and operational of Westinghouse AP1000 power reactor in several monitoring points at coolant reactor system. The results of the calculation of N-16 concentration at the core outlet, reactor vessel outlet, pressurizer line, inlet and outlet of steam generators, primary pumps, reactor vessels inlet and core inlet are: 281, 257, 255, 250, 145, 142, 129 and 112 µCi/gram respectively. The results of analysis compared with AP1000 design control document as standard values. The verification showed very high accuracy comparation between analytical results and standard values.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

Expanded Materials Degradation Analysis (EMDA) Volume 3: Aging of Reactor Pressure Vessels

In NUREG/CR-6923 ORNL/TM-2013/532, "Expert Panel Report on Proactive Materials Degradation Assessment," referred to as the PMDA report, NRC conducted a comprehensive evaluation of potential aging-related degradation modes for core internal components, as well as primary, secondary, and some tertiary piping systems, considering operation up to 40 years. This document has been a very valuable resource, supporting NRC staff evaluations of licensees' aging management programs and allowing for prioritization of research needs.This report describes an expanded materials degradation assessment (EMDA), which significantly broadens the scope of the PMDA report. The analytical timeframe is expanded to 80 years to encompass a potential second 20-year license-renewal operating-period, beyond the initial 40-year licensing term and a first 20-year license renewal. Further, a broader range of structures, systems, and components (SSCs) was evaluated, including core internals, piping systems, the reactor pressure vessel (RPV), electrical cables, and concrete and civil structures. The EMDA uses the approach of the phenomena identification and ranking table (PIRT), wherein an expert panel is convened to rank potential degradation scenarios according to their judgment of susceptibility and current state of knowledge.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Progress in the development of the neutron flux monitoring system of the French GEN-IV SFR: simulations and experimental validations [ANIMMA--2015-IO-392]

France has a long experience of about 50 years in designing, building and operating sodium-cooled fast reactors (SFR) such as RAPSODIE, PHENIX and SUPER PHENIX. Fast reactors feature the double capability of reducing nuclear waste and saving nuclear energy resources by burning actinides. Since this reactor type is one of those selected by the Generation IV International Forum, the French government asked, in the year 2006, CEA, namely the French Alternative Energies and Atomic Energy Commission, to lead the development of an innovative GEN-IV nuclear- fission power demonstrator. The major objective is to improve the safety and availability of an SFR. The neutron flux monitoring (NFM) system of any reactor must, in any situation, permit both reactivity control and power level monitoring from startup to full power. It also has to monitor possible changes in neutron flux distribution within the core region in order to prevent any local melting accident. The neutron detectors will have to be installed inside the reactor vessel because locations outside the vessel will suffer from severe disadvantages; radially the neutron shield that is also contained in the reactor vessel will cause unacceptable losses in neutron flux; below the core the presence of a core-catcher prevents from inserting neutron guides; and above the core the distance is too large to obtain decent neutron signals outside the vessel. Another important point is to limit the number of detectors placed in the vessel in order to alleviate their installation into the vessel. In this paper, we show that the architecture of the NFM system will rely on high-temperature fission chambers (HTFC) featuring wide-range flux monitoring capability. The definition of such a system is presented and the justifications of technological options are brought with the use of simulation and experimental results. Firstly, neutron-transport calculations allow us to propose two in-vessel regions, namely the above-core and under-core structures. We verify that they comply with the main objective, that is the neutron power and flux distribution monitoring. HTFC placed in these two regions can detect an inadvertent control rod withdrawal that is a postulated initiating event for safety demonstration. Secondly, we show that the HTFC reliability is enhanced thanks to a more robust physical design and the fact that it has been justified that the mineral insulation is insensitive to any increase in temperature. Indeed, the HTFC insulation is subject to partial discharges at high temperature when the electric field between their electrodes is greater than about 200 V/mm or so. These discharges give rise to signals similar to the neutron pulses generated by a fission chamber itself, which may bias the HTFC count rate at start-up only. However, as displayed in Figure 1, we have experimentally verified that one can discriminate neutron pulses from partial discharges using online estimation of pulse width. Thirdly, we propose to estimate the count rate of a HTFC using the third order cumulant of its signal that is described by a filtered Poisson process. For such a statistic process, it is known that any cumulant, also called cumulative moment, is proportional to the process intensity that is here the count rate of a fission chamber. One recalls that the so-called Campbelling mode of such a detector is actually based on the signal variance, which is the second-order cumulant as well. The use of this extended Campbelling mode based on the third-order cumulant will permit to ensure the HTFC response linearity over the entire neutron flux range using a signal processing technique that is simple enough to satisfy design constraints on electric devices important for nuclear safety. We also show that this technique, named high order Campbelling method (HOC), is significantly more robust than another technique based on the change in the HTFC filling gas, which consists in adding a few percent of nitrogen. Finally, we also present an experimental campaign devoted to the required calibration process of the so-called HOC method. The Campbelling results show a good agreement with the simple pulse counting estimation at low count rates. It is also shown that the HOC technique provides a linear estimation of the count rates at higher power levels as well.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Chooz A, First Pressurized Water Reactor to be Dismantled in France - 13445

Nine commercial nuclear power plants have been permanently shut down in France to date, of which the Chooz A plant underwent an extensive decommissioning and dismantling program. Chooz Nuclear Power Station is located in the municipality of Chooz, Ardennes region, in the northeast part of France. Chooz B1 and B2 are 1,500 megawatt electric (MWe) pressurized water reactors (PWRs) currently in operation. Chooz A, a 305 MWe PWR implanted in two caves within a hill, began operations in 1967 and closed in 1991, and will now become the first PWR in France to be fully dismantled. EDF CIDEN (Engineering Center for Dismantling and Environment) has awarded Westinghouse a contract for the dismantling of its Chooz A reactor vessel (RV). The project began in January 2010. Westinghouse is leading the project in a consortium with Nuvia France. The project scope includes overall project management, conditioning of the reactor vessel (RV) head, RV and RV internals segmentation, reactor nozzle cutting for lifting the RV out of the pit and seal it afterwards, dismantling of the RV thermal insulation, ALARA (As Low As Reasonably Achievable) forecast to ensure acceptable doses for the personnel, complementary vacuum cleaner to catch the chips during the segmentation work, needs and facilities, waste characterization and packaging, civil work modifications, licensing documentation. The RV and RV internals will be segmented based on the mechanical cutting technology that Westinghouse applied successfully for more than 13 years. The segmentation activities cover the cutting and packaging plan, tooling design and qualification, personnel training and site implementation. Since Chooz A is located inside two caves, the project will involve waste transportation from the reactor cave through long galleries to the waste buffer area. The project will end after the entire dismantling work is completed, and the waste storage is outside the caves and ready to be shipped either to the ANDRA (French National Radioactive Waste Management Agency) waste disposal facilities - (for low-level waste [LLW] and very low-level waste [VLLW], which are considered short lived) - or to the EDF Interim Storage Facility planned to be built on another site - (for low- and intermediate-level waste [LILW], which is considered long lived). The project has started with a detailed conceptual study that determines the step-by-step approach for dismantling the reactor and eventually supplying the packed containers ready for final disposal. All technical reports must be verified and approved by EDF and the French Nuclear Safety Authority before receiving the authorization to start the site work. The detailed conceptual study has been completed to date and equipment design and manufacturing is ongoing. This paper will present the conceptual design of the reactor internals segmentation and packaging process that will be implemented at Chooz A, including the planning, methodology, equipment, waste management, and packaging strategy. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Nuclear reactor assemblies, nuclear reactor target assemblies, and nuclear reactor methods

Reactor target assemblies are provided that can include a housing defining a perimeter of at least one volume and Np or Am spheres within the one volume. Reactor assemblies are provided that can include a reactor vessel and a bundle of target assemblies within the reactor vessel, at least one of the target assemblies comprising a housing defining a volume with Np or Am spheres being within the volume. Irradiation methods are also provided that can include irradiating Np or Am spheres within a nuclear reactor, then removing the irradiated spheres from the reactor and treating the irradiated spheres.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2011

HEAVY WATER COMPONENTS TEST REACTOR DECOMMISSIONING

The Heavy Water Components Test Reactor (HWCTR) Decommissioning Project was initiated in 2009 as a Comprehensive Environmental Response, Compensation and Liability Act (CERCLA) Removal Action with funding from the American Recovery and Reinvestment Act (ARRA). This paper summarizes the history prior to 2009, the major D&D activities, and final end state of the facility at completion of decommissioning in June 2011. The HWCTR facility was built in 1961, operated from 1962 to 1964, and is located in the northwest quadrant of the Savannah River Site (SRS) approximately three miles from the site boundary. The HWCTR was a pressurized heavy water test reactor used to develop candidate fuel designs for heavy water power reactors. In December of 1964, operations were terminated and the facility was placed in a standby condition as a result of the decision by the U.S. Atomic Energy Commission to redirect research and development work on heavy water power reactors to reactors cooled with organic materials. For about one year, site personnel maintained the facility in a standby status, and then retired the reactor in place. In the early 1990s, DOE began planning to decommission HWCTR. Yet, in the face of new budget constraints, DOE deferred dismantlement and placed HWCTR in an extended surveillance and maintenance mode. The doors of the reactor facility were welded shut to protect workers and discourage intruders. In 2009 the $1.6 billion allocation from the ARRA to SRS for site footprint reduction at SRS reopened the doors to HWCTR - this time for final decommissioning. Alternative studies concluded that the most environmentally safe, cost effective option for final decommissioning was to remove the reactor vessel, both steam generators, and all equipment above grade including the dome. The transfer coffin, originally above grade, was to be placed in the cavity vacated by the reactor vessel and the remaining below grade spaces would be grouted. Once all above equipment including the dome was removed, a concrete cover was to be placed over the remaining footprint and the groundwater monitored for an indefinite period to ensure compliance with environmental regulations.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Testing and Commissioning of a Multifunctional Tool for the Dismantling of the Activated Internals of the KNK Reactor Shaft - 13524

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 are already dismantled. The total contaminated sodium inventory has already been disposed of. Only the high activated reactor vessel shielding structures are remaining. Due to the high dose rates these structures must be dismantled remotely. For the dismantling of the primary shielding of the reactor vessel, 12 stacked cast iron blocks with a total mass of 90 Mg and single masses up to 15.5 Mg, a remote-controlled multifunctional dismantling device (HWZ) was designed, manufactured and tested in a mock-up. After successful approval of the test sequences by the authorities, the HWZ was implemented into the reactor building containment for final assembling of the auxiliary equipment and subsequent hot commissioning in 2012. Dismantling of the primary shielding blocks is scheduled for early 2013. (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 OSTI2017

Neutron Irradiation of Nuclear Structural Materials Fabricated by Powder Metallurgy with Hot Isostatic Pressing

An irradiation campaign intended to produce data necessary to qualify various reactor structural materials fabricated by powder metallurgy with hot isostatic pressing (PM/HIP) is currently in progress. Manufacturing processes have considerable influence over the safety and integrity of nuclear reactor vessels and internal components. Established processes such as casting, plate rolling and welding, forging, drawing, and extrusion have been used to fabricate structural and pressure-retaining materials used in the nuclear power industry for the past 60 years. However, weldability and inspectability are ongoing issues in the manufacture of reactor vessels and internals. Furthermore, these components are subject to harsh service environments, which combine high radiation fluence, high temperature, and mechanical stress. These conditions can accelerate material degradation and failure, with the most extreme degradation often occurring at weldments and in poorly manufactured or inspected components. Advanced reactor designs and lifetime extensions to the existing fleet of light-water reactors (LWRs) will exacerbate materials degradation issues by increasing the duty on reactor internals. Thus, developing reliable manufacturing processes wherein high-quality weldments and inspections can be performed is of great importance to the continued safety and operation of nuclear power plants.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Design study and comparative evaluation of JSFR failed fuel detection system

A conceptual design study of an advanced sodium-cooled fast reactor JSFR has progressed in the 'Fast Reactor Cycle Technology Development (FaCT) 'project in Japan. JSFR has two failed fuel detection systems in the core. One is a failed fuel detection (FFD) system which continuously monitors a fission product from failed fuel subassembly. The other is a failed fuel detection and location (FFDL) system which locates when it receives signals from FFD. The FFD system consists of a FFD-DN which detects delayed neutron (DN) in sodium and a FFD-CG which detects fission products in the cover gas of the reactor vessel. In this study, requirements to the FFD-DN and the FFD-DN design to meet the requirements were investigated for the commercial and demonstration JSFR. In the commercial JSFR, a sampling type FFD which collects sodium from the reactor vessel by sampling lines for DN detectors was adopted. The performances have been investigated and confirmed by a fluid analysis in the reactor upper plenum. In the demonstration JSFR, the performance of DN detectors installed on the primary cold-leg piping has been confirmed. For the FFDL systems, experiences in the previous fast reactors and the R and D of FFDL system for JSFR were investigated. This study focuses on the Selector-Valve and the Tagging-Gas FFDL systems. Operation experiences of the Selector-valve FFDL system were accumulated in PFR and Phenix. Tagging-gas system experiences were accumulated in EBR-II and FFTF. The feasibility of both FFDL systems for JSFR was evaluated. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Gripper Test Assembly Status Report (FY2020)

Refueling systems for fast reactors are designed to handle fresh and used core assemblies (fuel, reflector, and shield core assemblies) within the reactor vessel in an opaque coolant environment without visual reference. These refueling machines are designed to work in a sodium (or other fast reactor coolant) and argon vapor space environment and are engineered with the rotatable plug system to allow for the movement of fresh and spent fuel into and out of the reactor core. The refueling machines are a critical component in any reactor and thus need to undergo extensive testing in a prototypic environment to ensure that they will meet all of the system functions and requirements. Argonne has developed an innovative compact refueling system design for the Advanced Fast Reactor-100 that is based upon some mechanisms used in previous reactor designs, such as the U.K.’s Prototype Fast Reactor (PFR) and some mechanisms that have not been used in sodium. This compact refueling machine supports the reduction in size of the AFR-100’s reactor vessel, and if fully developed, would support and inform the development of the in-vessel refueling machines for such commercial reactors as the GEH PRISM reactor plant, the ARC Clean Energy’s ARC-100 reactor, and the Natrium reactor, among others. This refueling system is a vital component of a fast reactor that supports reducing the cost of the reactor and increasing its reliability. During the development of the compact fuel handling machine conceptual design for the AFR-100, a lack of testing data for many mechanical components in sodium under typical in-reactor loads and conditions was discovered. The reduction in lifetime of the various mechanical components in the liquid sodium environment needs to be quantified versus the calculated component lifetimes under normal conditions in the testing while they are subjected to typical loading profiles experienced in the past. The Gripper Test Assembly discussed here includes a full-size gripper device with appropriate mechanical features that will be tested in sodium to provide this testing data. The Gripper Test Assembly is used to test various mechanical fuel handling components submerged in high temperature liquid sodium. These mechanisms are gears, bearings, gripper jaws and head, universal joints and shafts, ball screws, among others. These components will be tested under the typical sodium environmental conditions experienced during refueling operations with appropriate loading conditions that simulate the removal and insertion of core assemblies into a fast reactor grid plate structure. This Gripper Test Assembly is the second in a series of refueling system mechanisms developed for testing in sodium. The first test assembly is the Gear Test Assembly (GTA) which was used to test the performance of gears and bearings operating in sodium. Because of the successful testing conducted with the GTA, it was decided to continue with the development of the Gripper Test Assembly which uses the same gears and bearings tested in GTA. Using the data collected during operation of this gripper test assembly, lifetime reduction factors of the various mechanical components can be calculated for the material combinations selected. These lifetime reduction factors can be used in the design of future mechanical systems which operate in these environments to accurately predict component end of life. In addition, an understanding of the ability of these components and mechanisms to operate under-sodium with the chosen materials will be accomplished. Again, this Gripper Test Assembly is the follow-on test article to the Gear Test Assembly which was testing the ability of gears and bearings operating in a sodium environment. Once fabricated and qualified, it will be tested in the Mechanism Engineering Test Loop (METL) facility located in Building 308 at Argonne National Laboratory.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Development and Experimental Validation of a Heat Transfer Model for Spilled Molten Salt Pools

A spill of radionuclide-bearing molten salt is one of the major postulated events that needs to be analyzed for liquid fluorine salt-cooled high-temperature reactor (FHR) or molten salt reactor licensing purposes. In this postulated event, radioactive source term materials (RSTMs) in the molten salt are discharged from the reactor vessel to the reactor building. The release of RSTMs from the spilled salt pool to the gas space in the reactor building is expected to be controlled by the cooling behavior of the spilled salt, including the growth and shrinkage of the solid crust on the surface of the spilled salt pool. This paper presents a simulation model for spilled salt pool heat transfer and validation efforts. The validation data come from two molten salt spill tests that were performed recently: the PELE2 test by the Rapid Experimental Laboratory of Kairos Power LLC (KP) and the Argonne salt cooling test conducted by Argonne National Laboratory. The former was a large-scale test involving kilograms of molten spilled FLiNaK salt, and the latter was a relatively smaller-scale test targeting various processes associated with a salt spill event. Both tests generated valuable data sets that can be used to assess salt cooling and validate evaluation models. This paper provides a new one-dimensional model that can simulate the cooling process of a spilled salt pool as well as the thermal responses of heat structures, such as the stainless steel liner and the concrete below the salt. The model has been implemented as part of KP-SAM code, which is a branch of the systems code SAM specific to KP FHR. In conclusion, the simulation results of the model are compared with the data of the PELE2 and Argonne tests, and reasonable agreements are observed between the model and test data.

heat transfer model↗
DOE OSTI2012

Application of the modified neutron source multiplication method to the prototype FBR Monju

The Modified Neutron Source Method (MNSM) is applied to the Monju reactor. This static method to estimate sub-criticality has already given good results on commercial Pressurized Water Reactors. The MNSM consists both in the extraction of the fundamental mode seen by a detector to avoid the effect of higher modes near sources, and the correction of flux distortion effects due to control rod movement. Among Monju's particularities that have a big influence on MNSM factors are: the presence of two californium sources and the position of the detector, which is located far from the core outside of the reactor vessel. The importance of spontaneous fission and ({alpha},n) reactions, which have increased during the shutdown period of 15 years, will also be discussed. In order to evaluate the detector count rate, an analytical propagation has been conducted from the reactor vessel. For two subcritical states, an estimation of the reactivity has been made and compared to experimental data obtained in the restart experiments at Monju (2010). Results indicate a good agreement between the MNSM reactivity and the reactivity measured with other methods. The reactivity dependence of the correction to apply to point kinetic equation is discussed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗