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At least 919 records · Page 51
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DOE OSTI2015

Analysis of fluid fuel flow to the neutron kinetics on molten salt reactor FUJI-12

Molten Salt Reactor is a reactor are operating with molten salt fuel flowing. This condition interpret that the neutron kinetics of this reactor is affected by the flow rate of the fuel. This research analyze effect by the alteration velocity of the fuel by MSR type Fuji-12, with fuel composition LiF-BeF{sub 2}-ThF{sub 4}-{sup 233}UF{sub 4} respectively 71.78%-16%-11.86%-0.36%. Calculation process in this study is performed numerically by SOR and finite difference method use C programming language. Data of reactivity, neutron flux, and the macroscopic fission cross section for calculation process obtain from SRAC-CITATION (Standard thermal Reactor Analysis Code) and JENDL-4.0 data library. SRAC system designed and developed by JAEA (Japan Atomic Energy Agency). This study aims to observe the effect of the velocity of fuel salt to the power generated from neutron precursors at fourth year of reactor operate (last critical condition) with number of multiplication effective; 1.0155.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

A Summary of the Department of Energy's Advanced Demonstration and Test Reactor Options Study

An assessment of advanced reactor technology options was conducted to provide a sound comparative technical context for future decisions by the Department of Energy (DOE) concerning these technologies. Strategic objectives were established that span a wide variety of important missions and advanced reactor technology needs were identified based on recent Department of Energy and international studies. A broad team of stakeholders from industry, academia and government was assembled to develop a comprehensive set of goals, criteria and metrics to evaluate advanced irradiation test and demonstration reactor concepts. Point designs of a select number of concepts were commissioned to provide a deeper technical basis for evaluation. The technology options were compared on the bases of technical readiness and the ability to meet the different strategic objectives. Using the study’s evaluation criteria and metrics, an independent group of exerts from industry, universities and national labs scored each of the point designs. Pathways to deployment for concepts of varying technical maturity were estimated for the different demonstration systems with regard to cost, schedule and possible licensing approaches. Furthermore, this study also presents the tradeoffs that exist among the different irradiation test reactor options in terms of the ability to conduct irradiations in support of advanced reactor research and development (R&D) and to serve potential secondary missions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

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↗
DOE OSTI2015

Nuclear Fuel Traces Definition in Storage Ponds of Research VVR-2 and OR Reactors in NRC 'Kurchatov Institute'

The Gas Plant complex is the experimental base of the Institute of Nuclear Reactors, which is part of the Kurchatov Institute. In 1954 the commissioning of the first Soviet water-cooled water-moderated research reactor VVR-2 on enriched uranium, and until 1983 the complex operated two research water-cooled water-moderated reactors 3 MW (VVR-2) and 300 kW (OR) capacity, which were dismantled in connection with the overall upgrades of the complex. The complex has three storage ponds in the reactor building. They are sub-surface vessels filled with water (the volume of water in each is about 6 m{sup 3}). In 2007-2013 the spent nuclear fuel from storages was removed for processing to 'Mayk'. Survey of Storage Ponds by Underwater Collimated Spectrometric System shows a considerable layer of slime on the bottom of ponds and traces of spent nuclear fuel in one of the storage. For determination qualitative and the quantitative composition of radionuclide we made complex α-, β-, γ- spectrometric research of water and bottom slimes from Gas Plant complex storage ponds. We found the spent nuclear fuel in water and bottom slime in all storage ponds. Specific activity of radionuclides in the bottom slime exceeded specific activity of radionuclides in the ponds water and was closed to levels of high radioactive waste. Analysis of the obtained data and data from earlier investigation of reactor MR storage ponds showed distinctions of specific activity of uranium and plutonium radionuclides. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Resourcing Strategies for the NRU Reactor Shutdown Planning and Execution - 19218

The National Research Universal Reactor (NRU) was commissioned in 1957. The NRU was fundamental for the development of the Candu reactors and, until 2016, quietly responsible for producing 40 percent of the world's supply of medical isotopes for diagnosis and cancer therapy. The Government of Canada decided in 2015 to discontinue operation as of March 31, 2018 due to challenges associated with running a 60-year old nuclear reactor. Anticipating this closing date allowed the Permanent Safe Shutdown Project (PSSP) to transition the NRU from an operating reactor to a state suitable for Storage with Surveillance. One of the biggest challenges to address was human resource management, which was rightly identified as the key to keeping the project within schedule and budget. The dilemma was, how to keep the work force motivated to contribute with the preparation of the shutdown while addressing job security concerns. This paper summarises Canadian Nuclear Laboratories (CNL) challenges and strategies to secure specialized subject matter experts, trades and operators for planning and executing the NRU reactor shutdown Project. The discussion also includes the solutions and lessons learned to date. (author)

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

Verification of Reactor Activation Modelling with Physical Characterization Data and the Impact on Long Term Assessments - 20303

Whiteshell Laboratories, located in Manitoba, Canada, provided research facilities for the Canadian nuclear industry since the early 1960's. The facility was centered on Whiteshell Reactor 1 (WR-1), an organically cooled, heavy water moderated nuclear reactor. WR-1 was safely shut down in 1985, defueled and drained, and has been in a safe Storage with Surveillance (Deferred Decommissioning) state until present day. In support of an Environmental Assessment for proposed in-situ disposal of WR-1, detailed characterization information regarding the reactor core components (calandria, calandria tubes, and pressure tubes) was necessary to provide confidence that the long-term impacts of the facility can be assessed adequately. A simplified estimate of the total activation of core components was completed in 1992 using conservative assumptions, expected reactor configuration and computer models built with WIMS-CRNL, ONEDANT and ORIGEN-S. Due to the simplification, and age of the work, there was considerable doubt that the modelled inventory estimate of the core components was sufficient to support long-term performance modeling of the planned in-situ decommissioning. Additional data would be required to provide this confidence. Several options were considered for characterizing the WR-1 core components, but it was decided to collect only a limited number of samples in an effort to validate the 1992 simplified estimate as bounding. The limited number of samples reduced risks to workers, both radiological and non-radiological, and significantly reduced the costs required to confidently define and bound the radiological inventory of activation products in the core. The results of the sampling were compared to the simplified activation estimation. There was good agreement between the modelled and sampled results, with all major contributors to total activity showing similar relative quantities. The samples were generally lower than predicted by the models, with few exceptions. Notable exceptions included Nb-94, which were higher in the sample results than the model. This was determined to be caused by the use of Niobium containing steel alloys in the stainless steel pressure tube that were not accounted for in the models. The comparison of results provided confidence that the original inventory estimates produced through the simplified model are conservative and bounding, and therefore suitable for use in the long-term assessment of in situ disposal of the WR-1 reactor. This paper provides details of the simplified computer model results, their comparison to results of the sampling, and the changes adopted in the assessment of the in-situ decommissioning of WR-1 as a result. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Recent research activities on high temperature pebble bed reactors at PSI - 18452

Recently, Paul Scherrer Institut (PSI) commenced a project on Feasibility and plausibility of innovative reactor concepts in an European electricity supply environment. The investigation in the project will be done on two reactor types: the pebble bed High-Temperature Gas-Cooled Reactor (HTGR) and Molten Salt Reactor (MSR) with the main focus being on the HTGR. Firstly, the project consists of an extensive literature study on the current status of the technology and research on the HTGR and MSR. Secondly, based on the findings from the literature survey, several topics of particular interest have been chosen for more thorough investigations. On the pebble bed HTGRs, the areas of investigation include risk assessment study, preliminary economic assessment, investigation of accident scenarios, and preliminary study on reduction of waste volumes. For both HTGR and MSR an analysis and optimization of the fuel cycle is foreseen. At the initial stage the equilibrium closed Th-U and U-Pu cycle performance will be evaluated and compared with 16 other reactor types. Later the knowledge from equilibrium study will be applied to optimize especially the MSR but also HTGR fuel cycle. At the end of the project, based on the literature survey and the results from the scoping studies, future research needs and areas of special interest will be identified for successive projects related to HTGR and MSR in PSI. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Experimental study on a simplified facility of HTGR reactor building response to depressurization accidental scenarios - 18516

The analysis and understanding of air ingress events is an important aspect in the design of The High Temperature Gas Cooled Reactor (HTGR) accident scenarios including Depressurized Loss of Forced Cooling (D-LOFC) events that allow for the possibility of air ingress into the Reactor Pressure Vessel (RPV) as the result of a break in the helium pressure boundary which can ultimately result in oxidation of the fuel elements and other nuclear-grade graphite components. In order to characterize the air ingress into the vented low pressure compartments (VLPC) of NGNP HTGR during hypothetical small break D-LOFC accidents experimentally, a 1/28 down scaled simplified reactor building model was developed. A non-dimensional similarity approach was employed for the scaling analysis of the experimental facility. Two experiments were conducted under similar experimental conditions, with 1.1 kg/h helium injection through the top of the reactor cavity, resulting in oxygen concentrations reaching to steady state within 4 minutes. Differences between the steady state oxygen concentrations results in the steam generator cavity of the two tests with the same experimental conditions were 0.51{+-}0.086% at the top of the steam generator cavity and 0.29{+-}0.11% at the bottom of the steam generator cavity. Difference in the pressure in the reactor between the two experiments were 0.09{+-}0.026 psig. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Design Report of the Molten-Chloride High Temperature Reactor (M-CHTR)

Liquid fueled molten salt reactors have seen renewed attention from reactor vendors and utility companies. In January of 2016, the Department of Energy awarded Southern Company, partnered with TerraPower, Electric Power Research Institute, Vanderbilt University, and Oak Ridge National Laboratory, a $40 million grant to support the development of the Molten Chloride Fast Reactor. This design report will discuss the design of the 2000 MW{sub th} Molten-Chloride High Temperature Reactor (M-CHTR). The M-CHTR is a liquid fuel molten sodium chloride fast reactor that is designed to consume spent nuclear fuel with normal operating temperatures near 1000 deg. C. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Transition to a U/Pu Fuel Cycle with Fast and Thermal Reactors

The purpose of this study was to analyze the transition from the current U.S. nuclear system to an advanced fuel cycle involving U/Pu continuous recycle in a fleet of fast and thermal reactors. This fuel cycle, known as 'EG29' in the 2014 Nuclear Fuel Cycle Evaluation and Screening (E and S), was identified as one of the four most promising out of a comprehensive list of fuel cycles that were grouped based on similar physics and performance characteristics. The EG29 fuel cycle involves continuously recycling the co-extracted U/Pu from a fleet of fast reactors represented by sodium-cooled fast reactors (SFRs) and burning the surplus amounts in a fleet of thermal reactors represented by pressurized water reactors with full mixed oxide (U,Pu)O{sub 2} cores (MOX-PWRs). The target energy generation ratio of SFRs to MOX-PWRs for the final EG29 system is 70:30. A 1% annual growth in nuclear energy demand was also assumed to be met during the entire scenario, including after EG29 is fully deployed. Therefore, to maintain system-wide mass balance, the SFR was designed to have a Pu inventory ratio (discharged/charged) of 1.4 compared to the MOX-PWR's 0.7. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Small CaH{sub 2} Moderated Thermal Reactor for Surface Power Generation I: Conceptual Design

National Aeronautics and Space Administration (NASA) published 'Journey to Mars' on October 8, 2015, stating their plan to send human to Mars for exploration, scientific research and colonization of Mars. Mission time on Martian surface is supposed to be ∼1 year, therefore, a surface power generation system is required for human habitation on Mars. Even though solar power and chemical battery have been used for the past missions, the former cannot generate power during month-scale dust storm on Mars, and the latter is not reliable for long-term use under a cold environment. On the other hand, nuclear power provides stable power even on Mars surface and, since kilowatt class power system is required in the planned mission, a fission power system is better than a radioisotope battery. An 8-year operation is required for the Mars surface power generation system with electrical power of 10 to 100 kWe and specific power of >5 We/kg, which means the system weight needs to be below 2000 kg in the case of 10 to 100 kWe system. In the current study, stretch goals for surface fission power system were set as follows: electric power of 20 kWe and total system weight below 1000 kg. In addition, nuclear security and critical safety are essential features for the surface fission power system considering launch accidents. A sufficient level of nuclear security is needed to prevent terrorists from obtaining nuclear materials even the reactor falls in a geographic area dominated by terrorists. From this view point, it is preferable not to use highly enriched uranium with a space reactor that can be diverted to nuclear weapons. Also, a nuclear reactor core should not reach criticality during a launch accident where the reactor core can be compressed by collision with sea or land. A novel concept of surface power reactor system that satisfies the above requirements is proposed in the current study. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Neutron Resonance Theory for Nuclear Reactor Applications: Modern Theory and Practices

The neutron resonance phenomena constitute one of the most fundamental subjects in nuclear physics as well as in reactor physics. It is the area where the concepts of nuclear interaction and the treatment of the neutronic balance in reactor fuel lattices become intertwined. The latter requires the detailed knowledge of resonance structures of many nuclides of practical interest to the development of nuclear energy. The most essential element in reactor physics is to provide an accurate account of the intricate balance between the neutrons produced by the fission process and neutrons lost due to the absorption process as well as those leaking out of the reactor system. The presence of resonance structures in many major nuclides obviously plays an important role in such processes. There has been a great deal of theoretical and practical interest in resonance reactions since Fermi’s discovery of resonance absorption of neutrons as they were slowed down in water. The resonance absorption became the center of attention when the question was raised as to the feasibility of the self-sustaining chain reaction in a natural uranium-fueled system. The threshold of the nuclear era was crossed almost eighty years ago when Fermi and Szilard observed that a substantial reduction in resonance absorption is possible if the uranium was made into the form of lumps instead of a homogeneous mixture with water. In the West, the first practical method for estimating the resonance escape probability in a reactor cell was pioneered by Wigner et al in early forties. (author)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Transient analysis of a micro-reactor using the DireWolf code suite

Transient analyses of heat pipe micro-reactors are necessary to ensure that hypothetical accident scenarios do not comprise reactor safety. Due to its small size and reliance on heat-pipes for cooling, the micro-reactor design introduced in this paper is a tightly coupled system which requires multi-physics tools to accurately model transient events. Idaho National Laboratory's DireWolf code suite based on the MOOSE framework is tailor-built to model heat-pipe reactors. This paper demonstrates DireWolf's ability to simulate the coupled thermal-neutronics transient behavior of a heat-pipe micro-reactor. The transient events presented here include an inadvertent rotation of all control drums simultaneously and a sudden complete rotation of a single control drum. A detailed description of each event is provided along with simulation results, including time dependent power and temperature distributions, and discussion. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Fast modular reactor nuclear design parameters of fuel cycle and power distributions

This paper describes conceptual design of a 112 MW thermal (50 MW electric) Fast Modular Reactor (FMR) system operating at 7 MPa with inlet/outlet temperatures of 509/800 Celsius degrees. The reactor system includes the reactor core, fuel assemblies, fuel rods, reactor internals, reflector, neutron control system, flow control, and structural components. The nuclear design and analysis were conducted to search for a baseline core with a cycle length greater than 8-year and the power peaking factor less than 1.5 by adjusting the fuel assembly, reactor core, and reflector configurations. The neutron transport calculations of the baseline core showed that a refueling interval of 9-year is achieved with a total peaking factor of 1.47 and a fuel rod average linear power of 3.6 kW/m. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Development of a high fidelity model of the CROCUS experimental reactor

Measurements of scalar flux distributions with fine spatial and energy resolutions are needed to remedy one of the validation shortcomings of the novel neutronics full core solvers, such as MPACT and nTRACER. While a very detailed resolution of the flux can be calculated with such codes, only a limited experimental data set is available to check their accuracy. Such type of measurements are on-going at the zero power reactor CROCUS, operated at the Laboratory for Reactor Physics and System Behaviours of the EPFL, thanks to the development of advanced miniature neutron detection systems. This kind of experimental data would provide the community with a suitable benchmark for the validation of high fidelity neutronics solvers. In parallel, a multi-physics solver for steady-state and transient analysis of nuclear reactors, named GeN-Foam, has been developed. Based on the finite-volume OpenFOAM library, GeN-Foam provides us with enough flexibility to analyze non-conventional reactor geometries such as that of CROCUS. While CROCUS heterogeneities cannot be modeled by MPACT and nTRACER for the moment, GeN-Foam offers a unique opportunity to build a high fidelity model which mimics these codes' method to reach sub-pin simulation resolution. This document aims at describing the work achieved to get from the existing GeN-Foam model of the CROCUS reactor based on a structured mesh and using the neutron diffusion, the first high-fidelity model using discrete ordinates method as an approximation to neutron transport and an unstructured mesh for inter lattice water gap description and sub-pin heterogeneous modeling. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

A hybrid porous model for full reactor core scale CFD investigation of a prismatic HTGR

Three-dimensional (3-D) Computational Fluid Dynamics (CFD) analysis of a whole nuclear reactor core is a tremendous challenge due to the large geometric volume and complex structures. Here, this research presents a hybrid porous (HP) model to simplify a prismatic High Temperature Gas-cooled Reactor (HTGR) core, so 3-D CFD investigation can be performed on a full reactor core scale. In the HP model, the prototypic small coolant channels in the nuclear fuel blocks are lumped together to form multiple equivalent large coolant channels, and then the porous medium flow model is applied to each of them. Therefore, heat transfer in fuel blocks is computed by a hybrid combination of solid energy and porous flow energy equations. The similarity between the HP model and prototypic model is achieved by deriving the porous flow permeability, inertial resistance factor, and artificial thermophysical properties. Compared with the widely used whole porous (WP) flow model, the HP model preserves more realistic geometric structures, and therefore more accurate physical processes. The General Atomics' Modular High Temperature Gas-cooled Reactor (MHTGR) design was chosen as a prototype to demonstrate the methodology. Simulations were performed using the prototypic CFD model and HP model at steady-state forced circulation, steady-state natural circulation, and transient conditions that correspond to normal operation, extended period of pressurized cool down, and short-term transients after reactor shutdown, respectively. The comparison shows good agreement between the HP model and prototypic model in the maximum fuel temperature, average solid temperature, and helium flow rate, which demonstrates the potential applicability of the HP model for a full reactor core scale simulation in the future. As a benefit, the HP model reduces the mesh quantity by a factor of 50 from a prototypic model. Correspondingly, the computation time was reduced by a factor of at least 30.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
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 OSTI2022

Two-step neutronics calculations with Shift and Griffin for advanced reactor systems

This research develops the initial coupling of the Shift Monte Carlo (MC) code and the Griffin reactor physics code for reactor analysis of non–light-water reactor systems. The novelty of this work is twofold. It is the first application of Shift to produce the multigroup cross sections needed for Griffin as applied to a non–light-water reactor system; and, the first investigation and analysis of characteristics of the Empire microreactor benchmark that should be considered for steady state and transient reactor physics calculations. This application uses the previously developed two-step neutronics analysis workflow to demonstrate this initial coupling. Here, we outline the two-step neutronics analysis workflow in which the Shift MC code is used to generate the multigroup cross sections and fluxes needed by the Griffin deterministic solver. Details on how these multigroup cross sections are generated using MC tallies are given, as well as the practicalities and limitations of the two-step neutronics workflow. The Empire microreactor benchmark was used to investigate and validate this coupling. Results using this benchmark show good agreement between Griffin calculations using Serpent-generated cross sections and Shift-generated cross sections. Analysis of the characteristics of this Empire benchmark show larger eigenvalue differences between heterogeneous and pin–homogenized solutions compared to those of traditional light-water reactor (LWR) designs, thus requiring super homogenization factor corrections for accurate eigenvalue and power distribution predictions.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗