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At least 145 records · Page 8
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Establishing a Database of High Temperature Salt Properties for Nuclear Applications

The Fluoride-Salt-Cooled High-Temperature Reactor (FHR) is a new reactor concept employing a high temperature liquid salt fluoride coolant. The thermophysical property data for candidate salt coolants is very limited. The uncertainties of these properties are quite high and not well documented. For researchers exploring reactor designs such as the FHR, a database of high temperature liquid salts for nuclear applications is being established. Researchers can use the database, which will be updated with new experimental data, to find thermophysical properties and associated uncertainties for salt coolants to use in their analyses. This paper serves as an introduction of the database and highlights the main three coolants in the database that are considered for use in the FHR design. The salts considered for the Fluoride-Salt-Cooled High- Temperature Reactor (FHR) include LiF-BeF{sub 2} (flibe) and NaFZrF{sub 4} (nafzirf) as potential primary loop coolants and LiF-NaFKF (flinak) as a secondary loop coolant. Flibe is the primary candidate for the primary loop coolant. It has the lowest neutron cross section, is chemically stable with low corrosion rates in high-nickel alloy systems, and is weakly radioactive with the primary activation product tritium resulting in very low radiation levels in the primary loop. Flibe does have some disadvantages, which include toxicity of beryllium and the need to enrich lithium to at least 99.995% Li-7 to maintain a low nuclear cross section. For design analysis, the baseline salt used is nafzirf because it has less attractive neutronic and thermophysical properties. The benefits of nafzirf over flibe including no direct tritium production and lower cost. By designing a reactor to operate using nafzirf, it is assumed that the reactor will operate better using flibe as the coolant. Thus, one can consider the coolants flibe and nafzirf to be bounding primary salt coolants for reactor design. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Development of Heat Pipe Reactor Modeling in SAM

System Analysis Module (SAM) is under development at Argonne National Laboratory as a modern system-level modeling and simulation tool for advanced non-light water reactor safety analyses. It utilizes the object-oriented application framework MOOSE to leverage the modern software environment and advanced numerical methods available in PETSc. The capabilities of SAM are being extended to enable the transient modeling, analysis, and design of various advanced nuclear reactor systems. This report presents the development of new capabilities for modeling the heat pipe type reactor systems. The need for power at remote locations away from a reliable electrical grid is an important niche for nuclear energy. Heat pipe-cooled fast-spectrum nuclear reactors are well suited for these applications. The key feature of the heat pipe reactors is the use of heat pipes for heat removal from the reactor core. The heat pipe makes use of the phase change of the working fluid and transports a large amount of heat from the evaporator to the condensation end with very small temperature drops. In contrast to the traditional nuclear reactor system that makes use of pumped loop for extracting the thermal power, the heat pipe reactors make use of hundreds of heat pipes for removing the thermal power (including the decay heat) passively. This could potentially significantly improve the reliability and safety of the reactor systems. The essential part in the analysis of a heat pipe type reactor is the modeling of heat transport inside the heat pipe. The capability of SAM is extended in this work to enable the modeling of the conventional heat pipe. Two alternative modeling options, 2D-RZ Heat conduction and 3D-1D coupling, are developed, which are both based on the assumption that the heat pipe vapor core can be modeled as a superconductor of extremely high thermal conductivity. Both modeling options are verified with a simplified thermal resistance model. The heat pipes are also coupled with the fuel region of a prototype micro reactor under both normal and off-normal conditions. It is confirmed that both modeling options can correctly model the heat transport in the heat-pipe reactors.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Modernization of Technical Requirements for Licensing of Advanced Non-Light Water Reactors: Selection and Evaluation of Licensing Basis Events

This report, “Modernization of Technical Requirements for Licensing of Advanced Non-Light Water Reactors: Selection and Evaluation of Licensing Basis Events,” represents a key element in the development of a methodology for the efficient licensing of advanced non-light water reactors (non-LWRs).

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Modernization of Technical Requirements for Licensing of Advanced Non-Light Water Reactors: Safety Classification and Performance Criteria for Structures, Systems, and Components

This report, “Modernization of Technical Requirements for Licensing of Advanced Non-Light Water Reactors: Safety Classification and Performance Criteria for Structures, Systems and Components,” represents a key element in the development of a framework for the efficient licensing of advanced non-Light Water Reactors (non-LWRs).

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

High-sensitivity fast neutron detector KNK-2-7M

The construction of the fast neutron detector KNK-2-7M is briefly described. The results of the study of the detector in the pulse-counting mode are given for the fissions of {sup 237}Np nuclei in the radiator of the neutron-sensitive section and in the current mode with the separation of sectional currents of functional sections. The possibilities of determining the effective number of {sup 237}Np nuclei in the radiator of the neutronsensitive section are considered. The diagnostic possibilities of the detector in the counting mode are shown by example of the analysis of the reference data from the neutron-field characteristics in the working hall of the BR-K1 reactor. The diagnostic possibilities of the detector in the current operating mode are shown by example of the results of measuring the {sup 237}Np-fission intensity in the BR-K1 reactor power start-ups implemented in the mode of fission-pulse generation on delayed neutrons at the detector arrangement inside the reactor core cavity under conditions of a wide variation of the reactor radiation field.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Corrosion performance of advanced structural materials in sodium.

This report gives a description of the activities in design, fabrication, construction, and assembling of a pumped sodium loop for the sodium compatibility studies on advanced structural materials. The work is the Argonne National Laboratory (ANL) portion of the effort on the work project entitled, 'Sodium Compatibility of Advanced Fast Reactor Materials,' and is a part of Advanced Materials Development within the Reactor Campaign. The objective of this project is to develop information on sodium corrosion compatibility of advanced materials being considered for sodium reactor applications. This report gives the status of the sodium pumped loop at Argonne National Laboratory, the specimen details, and the technical approach to evaluate the sodium compatibility of advanced structural alloys. This report is a deliverable from ANL in FY2010 (M2GAN10SF050302) under the work package G-AN10SF0503 'Sodium Compatibility of Advanced Fast Reactor Materials.'

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Sensitivity Analysis of an Experimental Breeder Reactor II Fuel Assembly

The Experimental Breeder Reactor II (EBR-II) was selected as a candidate for a reactor physics benchmark evaluation due to its unique features and capabilities. In connection with the benchmark evaluation, the goal of this work was to perform a sensitivity analysis. The EBR-II was a sodium cooled, pool type, metal fueled, fast spectrum reactor. EBR-II operated from 1964 through 1994. It had a maximum heat output of 62.5 MW which could produce 20 MW of electricity. The design of EBR-II consisted of 637 hexagonal shaped, removable assemblies. The assemblies were divided into three regions: the core, an inner blanket, and an outer blanket. The core region contained assemblies with 91 fuel rods in each. The fuel was enriched uranium metal and was clad within stainless steel. The core region also included two safety assemblies and eight control assemblies with the difference being that these assemblies only contained 61 fuel rods. Additional assemblies in the core region included stainless steel dummies, half worth drivers, and experimental/instrumentation assemblies. The inner blanket region initially consisted of depleted uranium for fuel breeding, but was replaced, after proving the concept of breeding, with stainless steel reflectors for the purpose of irradiation. The outer blanket region consisted of depleted uranium. Figure 1 shows a cross section view of the EBR-II core layout. Each assembly was approximately 234 cm long with the uranium fuel section being 33 cm long. Above and below the fuel were stainless steel neutron reflectors. Additionally, orifices at the bottom of the assembly allowed for sodium coolant to flow upward, with larger holes in the subassemblies at the center where the greatest heat was produced. The core of EBR-II was enclosed by a stainless steel reactor vessel which was comprised of a radial shell, a lower grid plenum structure, and an upper reactor vessel cover that served as a neutron shield. Surrounding the reactor vessel radially were layers of graphite and borated graphite shielding. The entire assembly is submerged 3 meters in a pool of sodium. The pool contained up to 337,000 liters of sodium which was forced through the core and blanket regions using two centrifugal pumps that were also submerged within the sodium pool. The design of EBR-II included many safety features. The compatibility between the sodium and stainless steel prevented corrosion and therefore the risk of a radioactive release due to corrosion. The sodium coolant was kept at near atmospheric pressure which made any leak more controllable. The metal fuel, cladding, and coolant allowed for heat to be removed easily. The large volume of sodium allowed for heat to be absorbed without problems occurring. Lastly, the reactor would expand as temperature increased which would result in greater neutron leakage. This extra leakage caused a large negative temperature reactivity feedback which automatically shut the reactor down in case of an accident. EBR-II had unique capabilities which were proven during the Integral Fast Reactor initiative. This program lasted from 1984 to 1994. The goal of this program was to solve many of the perceived problems with nuclear energy through scientific means. The safety features of EBR-II were put to the test in April 1986. The tests simulated loss of reactor coolant flow and loss of a heat sink. For both of these tests, the reactor was at 100% power and was not scrammed. During both tests, the reactor shut itself down and achieved a safe steady state temperature all due to the design of the reactor. It was because of these tests that EBR-II was chosen for a reactor physics benchmark evaluation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRSDate not supplied

Development of a Cold-Walled Molten Regolith Electrolysis Reactor for Lunar Oxygen Production

On the lunar surface, production of commodi-ties to support human presence, such as water, food and oxygen, and sustain the growth of a per-manent outpostwill likely require the use oflocal resources. The moon is covered almost entirely withfragmented oxide minerals known as regolithhundreds of meter thick.As a resource, it is rich in oxygen (> 42 wt.%) bound in a solid state with a variety of metals. The molten regolith electrolysis (MRE)reactor is a promising technology for the production of gaseous oxygen from the lunar reg-olithin a simple, single-stepreaction that requires minimalconsumable materials, produces oxygen and metals with high electrical efficiency and high yields from any regolithcomposition.This process involvesmelting regolith to~1600°C then electro-lyzing the molten pool to separate metal and oxy-gen ions that are then collected as liquid metal and gaseous oxygen at the respective electrodes. Lab-scale demonstrations of the MRE technology have previously reliedon external heating sources to bring the entirety of the reactor up to the operating temperaturewhich creates corrosive interfaces be-tween the molten regolith and the containment ma-terial in the reactor, limiting the overall lifespan of a reactor[1]. The Gaseous Lunar Oxygen from Regolith Electrolysis (GaLORE) project is focused on the development of a “cold-walled” or “Joule-heated” reactor design in which an internal heating source is used to selectively melt a pool of regolith between the electrodes of the reactor, leaving a shell of solidified regolith between the molten pool and the containment vessel of the reactor. This next generation reactor concept has been under development as molten oxide electrolysis (MOE) by MIT and Boston Metal for the production of iron from pure ores for terrestrial application [2]. The GaLORE project in engaged in early development of the technology for use with varying lunar regolith compositionsin the lunar environment. Thermal modelling of a proposed cold-walledreac-tor design were used as a scaffold to develop pa-rameters for a feasible reactor shape and size as well as target energy consumption[3]. The current development effort for the cold-walled reactor de-sign will be presented as a trade study of the most promising techniques for melting regolithwithin the constraints imposed by the lunar environment.Heater devices are designed to accommodate lim-ited electrical power availabilityon the moon, a wide range of regolith compositions that may be seen on the moon, limited metals available for re-placing consumed parts,and the low thermal con-ductivity of granular regolith in vacuum.Heater de-vices will be down-selectedbased on performance measurements within the above operationalcon-straints,and selected devices will be integrated into a reactor with electrodes to begin producing oxygen. [1]Sibille,L.,Sadoway, D.R.,Sirk, A., Tripathy,P.,Melendez, O., Standish, E., Dominguez, J. A., Stefanescu, D.M.,Curreri, P.A., Poizeau,S.,2009. “Recent Advancesin Scale-up Development of Molten Regolith Electrolysis for Oxygen Production in support of a Lunar Base.”AIAA 2009-659, 47th AIAA Aerospace Sciences Meeting, 5 -8 January 2009, Orlando, FL. [2] Boston Metal, https://www.boston-metal.com/moe-technology/#moe-process[3]Schreiner, S.S.,Sibille, L., Dominguez, J.A., Hoffman, J.A., 2016. "A parametric sizing model for Molten Regolith Electrolysis reactors to pro-duce oxygen on the Moon." Advances in Space Research 57.,7,1585-1603.

K D Grossman↗
DOE OSTI2016

Light Water Reactor Sustainability Program Reactor Safety Technologies Pathway Technical Program Plan

In the aftermath of the March 2011 multi-unit accident at the Fukushima Daiichi nuclear power plant (Fukushima), the nuclear community has been reassessing certain safety assumptions about nuclear reactor plant design, operations and emergency actions, particularly with respect to extreme events that might occur and that are beyond each plant’s current design basis. Because of our significant domestic investment in nuclear reactor technology (99 operating reactors in the fleet of commercial LWRs with five under construction), the United States has been a major leader internationally in these activities. The U.S. nuclear industry is voluntarily pursuing a number of additional safety initiatives. The NRC continues to evaluate and, where deemed appropriate, establish new requirements for ensuring adequate protection of public health and safety in the occurrence of low probability events at nuclear plants; (e.g., mitigation strategies for beyond design basis events initiated by external events like seismic or flooding initiators). The DOE has also played a major role in the U.S. response to the Fukushima accident. Initially, DOE worked with the Japanese and the international community to help develop a more complete understanding of the Fukushima accident progression and its consequences, and to respond to various safety concerns emerging from uncertainties about the nature of and the effects from the accident. DOE R&D activities are focused on providing scientific and technical insights, data, analyses methods that ultimately support industry efforts to enhance safety. These activities are expected to further enhance the safety performance of currently operating U.S. nuclear power plants as well as better characterize the safety performance of future U.S. plants. In pursuing this area of R&D, DOE recognizes that the commercial nuclear industry is ultimately responsible for the safe operation of licensed nuclear facilities. As such, industry is considered the primary “end user” of the results from this DOE-sponsored work. The response to the Fukushima accident has been global, and there is a continuing multinational interest in collaborations to better quantify accident consequences and to incorporate lessons learned from the accident. DOE will continue to seek opportunities to facilitate collaborations that are of value to the U.S. industry, particularly where the collaboration provides access to vital data from the accident or otherwise supports or leverages other important R&D work. The purpose of the Reactor Safety Technology R&D is to improve understanding of beyond design basis events and reduce uncertainty in severe accident progression, phenomenology, and outcomes using existing analytical codes and information gleaned from severe accidents, in particular the Fukushima Daiichi events. This information will be used to aid in developing mitigating strategies and improving severe accident management guidelines for the current light water reactor fleet.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

An integrated microbial electrolysis-anaerobic digestion process combined with pretreatment of wastewater solids to improve hydrogen production

A combined anaerobic digestion (AD) and microbial electrolysis cell (MEC) system, named here as ADMEC, was investigated in this paper to evaluate the energy recovery from pretreated wastewater solids. Alkaline and thermal hydrolysis pretreatment methods increased the solubility of organic compounds present in the raw solids by 25% and 20%, respectively. The soluble phase from pretreatment was separated and used for microbial electrolysis, whereas the insoluble fraction was fed into semi-continuous digesters. The digester effluent was later utilized as a second MEC substrate. The pretreatment had variable effects on AD and MEC performance. The methane content in AD biogas was higher in pretreated groups, 78.29 ± 2.89% and 73.2 ± 1.79%, for alkaline and thermal, than the control, 50.26 ± 0.53%, but the overall biogas production rates were lower than the control, 20 and 30 mL CH 4 gCOD -1 d -1 for alkaline and thermal compared to 80 mL CH 4 gCOD -1 d -1 . The effluent streams from thermally pretreated digesters were the best substrate for microbial electrolysis, in terms of hydrogen production and efficiency. The MECs produced 1.7 ± 0.2 L-H 2 per L per day, 0.3 ± 0.1 L-H 2 per L per day, and 0.29 ± 0.1 L-H 2 per L per day, for thermal, alkaline, and control reactors. The productivity was lower compared to acetate and propionate controls, which yielded 5.79 ± 0.03 L-H 2 per L per day and 3.49 ± 0.10 L-H 2 per L per day, respectively. The pretreatment solubilized fractions were not ideal substrates for microbial electrolysis. Finally, a chemical oxygen demand (COD) mass balance showed that pretreatment shifts the electron flux away from methane and biomass sinks towards hydrogen production.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2019

Design of an experimental breeder reactor run 138B reactor physics benchmark evaluation management application

Reactor Monte Carlo simulations present numerous challenges including input file generation and processing output files for relevant results. Creating an input file for a heterogeneous core is further complicated when using depleted fuel. Performing perturbations to geometry and material characteristics exacerbates the challenge due to human errors if the input files are manually generated. The difficulty of generating input files is often encountered during the process of performing a reactor physics benchmark evaluation. These challenges were encountered in this study when performing a reactor physics benchmark evaluation for the Experimental Breeder Reactor II (EBR-II), a sodium cooled fast reactor. The EBR-II Monte Carlo input file uses more than 1,000 unique material specifications and ~8,000 regions. To address the input file generation challenge, a computer application was designed to automate the input file generation process. The application utilizes user input from a graphical user interface in conjunction with spreadsheets containing material and geometric data to generate Monte Carlo input files. The application, called MICKA, was used to generate input files to support the reactor physics benchmark evaluation for EBR-II. This paper describes the reactor configuration and provides information about the application design and capabilities for addressing challenges often encountered in reactor physics benchmark evaluations.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Environmental creep-fatigue and weld creep cracking: a summary of design and fitness-for-service practices

This report surveys current creep-fatigue design and fitness-for-service assessment procedures, with a particular focus on environmental effects and creep-fatigue cracking near weldments. Included in this assessment are the American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section III, Division 5, the British R5 standard, the French RCC-MRx design code, the ITER Structural Design Criteria, and the API-579/ASME FFS-1 fitness-for-service rules for petrochemical components. The focus of this review is to identify gaps in the available methods that could prevent regulators from evaluating future high temperature reactor designs. The report makes recommendations, where appropriate, on actions that could be taken to fill these gaps by either providing adequate assessment procedures for creep-fatigue damage in harsh environments or by evaluating strategies vendors could use to mitigate regulatory concerns with creep-fatigue damage in high temperature reactor components.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Development of a diversity and defense-in-depth strategy for the CNNC Fuqing, Fangjiashan and Hainan nuclear plants - 379

The purpose of this paper is to provide an overview of the development of the Diversity and Defense-in- Depth (D3) strategy for the China National Nuclear Corporation Fuqing, Fangjiashan, and Hainan Nuclear Plants. The Nuclear Power Plant (NPP) Digital Control System (DCS) and advanced control room was completed by Invensys and China Nuclear Power Engineering (CNPE)/Nuclear Power Institute of China (NPIC) for these eight new Chinese built Pressurized Water Reactor plants. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

High-Sensitivity Fast Neutron Detector KNK-2-8M

The design of the fast neutron detector KNK-2-8M is outlined. The results of he detector study in the pulse counting mode with pulses from {sup 238}U nuclei fission in the radiator of the neutron-sensitive section and in the current mode with separation of functional section currents are presented. The possibilities of determination of the effective number of {sup 238}U nuclei in the radiator of the neutron-sensitive section are considered. The diagnostic capabilities of the detector in the counting mode are demonstrated, as exemplified by the analysis of reference data on characteristics of neutron fields in the BR-1 reactor hall. The diagnostic capabilities of the detector in the current mode are demonstrated, as exemplified by the results of measurements of {sup 238}U fission intensity in the power startup of the BR-K1 reactor in the fission pulse generation mode with delayed neutrons and the detector placed in the reactor cavity in conditions of large-scale variation of the reactor radiation fields.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI1991

Nuclear Safety [Vol. 32, No. 4, October-December 1991]

Nuclear Safety is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 477 Report on the International Symposium on the Use of Probabilistic Safety Assessment for Operational Safety—PSA '91, S. Chakraborty and M. Khatib-Rahbar; 488 Good Relationships Are Pivotal in Nuclear Data Bases, A. S. Heger and B. V. Koen; 494 Technical Note: The Interagency Nuclear Safety Review Panel's Evaluation of the Ulysses Space Mission, J. A. Sholtis, Jr., D. A. Huff, L. B. Gray, N. P. Klug, and R. O. Winchester; ACCIDENT ANALYSIS: 502 The Severe Accident Analysis Program for the Savannah River Nuclear Production Reactors, M. L. Hyder; CONTROL AND INSTRUMENTATION: 511 A Framework for Selecting Suitable Control Technologies for Nuclear Power Plant Systems, R. A. Kisner; DESIGN FEATURES: 521 Containments for Gas-Cooled Power Reactors History and Status, P.W. Williams; ENVIRONMENTAL EFFECTS: 537 Indoor Radon: A Natural Risk, N. H. Harley and J.H. Harley; On the Importance of the Atmospheric Parameters in the Fission Products Distribution of a Severe Reactor Accident, M. C. Barla and A. R. Bayulken; Book Review: Health Effects of Exposure to Low Levels of Ionizing Radiation BEIR V, C. R. Richmond; WASTE AND SPENT FUEL MANAGEMENT: 555 Activities Related to Waste and Spent Fuel Management, M. D. Muhlheim and E. G. Silver OPERATING EXPERIENCES: 567 Effects of Component Aging on the Westhinghouse Control Rod Drive System, K. Sullivan and W. Gunther; 577 Reactor Shutdown Experience, Compiled by J. W. Cletcher, 580 Selected Safety-Related Events, Compiled by G. A. Murphy; 582 Operating U 8 Power Reactors, Compiled by M. D. Muhlheim and E. G. Silver, RECENT DEVELOPMENTS: 596 General Administrative Activities, Compiled by M. D. Muhlheim and E. G. Silver; 610 Reports, Standards, and Safety Guides, D. S. Queener; 614 Proposed Rule Changes as of June 30, 1991; ANNOUNCEMENTS: 520 Workshop on PC PRAISE: A Probabilistic Fracture Mechanics Personal Computer Code for Nuclear Power Plant Piping Reliability Assessment; 536 Fifth Workshop on Nuclear Power Plant Containment Integrity; 624 New OECD "International Information System on Occupational Exposure (ISOE)"; 625 Harvard School of Public Health Announces Short Courses; 625 Eighth Power Plant Dynamics, Control and Testing Symposium; 626 Second Training Course on Off-Site Emergency Planning and Response for Nuclear Accidents; 618 The Authors; 622 Reviewers of Nuclear Safety, Vol 32.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗