Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “L REACTOR”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 937 records · Page 52
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

Haven't a Cue? Mapping the CUE Space as an Aid to HRA Modeling

Advances in automation present a new modeling environment for the human reliability analysis (HRA) practitioner. Many, if not most, current day HRA methods have their origin in characterizing and quantifying human performance in analog environments where mode awareness and system status indications are potentially less comprehensive, but simpler to comprehend at a glance when compared to advanced presentation systems. The introduction of highly complex automation has the potential to lead to: decreased levels of situation awareness caused by the need for increased monitoring; confusion regarding the often non-obvious causes of automation failures, and emergent system dependencies that formerly may have been uncharacterized. Understanding the relation of incoming cues available to operators during plant upset conditions, in conjunction with operating procedures, yields insight into understanding the nature of the expected operator response in this control room environment. Static systems methods such as fault trees do not contain the appropriate temporal information or necessarily specify the relationship among cues leading to operator response. In this paper, we do not attempt to replace standard performance shaping factors commonly used in HRA nor offer a new HRA method, existing methods may suffice. In this paper we strive to enhance current understanding of the basis for operator response through a technique that can be used during the qualitative portion of the HRA analysis process. The CUE map is a means to visualize the relationship among salient cues in the control room that help influence operator response, show how the cognitive map of the operator changes as information is gained or lost, and is applicable to existing as well as advanced hybrid plants and small modular reactor designs. A brief application involving loss of condensate is presented and advantages and limitations of the modeling approach and use of the CUE map are discussed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

First experimental tests of a new small angle slot divertor on DIII-D

A new small angle slot (SAS) divertor concept has been developed to enhance neutral cooling across the divertor target by coupling a closed slot structure with appropriate target shaping. Initial tests on DIII-D find a strong interplay between such anticipated ‘SAS’ effects and cross-field drifts, favouring operation with the ion B×∇B drift away from the X-point, as currently employed for advanced tokamaks. This offers the following key improvements relative to DIII-D’s open lower divertor or partially-closed upper divertor: (i) SAS allows for transition to low temperature moderately detached divertor conditions with T_e≲10eV at very low main plasma densities, lower than are usually attainable at all in DIII-D high confinement (H-mode) plasmas as used in these tests; (ii) Pedestal performance and core confinement are significantly improved with SAS. The final confinement collapse associated with the onset of X-point MARFE (multifaceted asymmetric radiation from the edge) following deep detachment occurs at significantly higher pedestal densities, thus widening the window of H-mode operation compatible with a dissipative divertor. For operation with the ion B×∇B drift toward the X-point, the divertor plasma transitions to a bifurcative detached state at much higher densities, similar to other divertor configurations in DIII-D. These results highlight the strong interplay between divertor closure and drifts, and point to an interesting divertor optimization path to explore that offers potential for future fusion reactors.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Incorporation of Thermal Hydraulic Models for Thermal Power Dispatch into a PWR Power Plant Simulator

This report describes the development, modeling, and results of a generic pressurized water reactor power plant simulator that incorporates coupled electrical and thermal power dispatch to an industrial process located approximately one kilometer from the nuclear power plant. The simulator is a commercial PWR simulator that has been modified to include thermal power dispatch as described in past milestone reports [ , ]. The commercial PWR simulator is a generic simulator available from GSE SYSTEMS® (Sykesville, MD, USA) that is built using RELAP5-HDTM Real-Time Solution and in-house software developed by GSE Systems. This generic PWR (GPWR) simulator performs real-time simulation of the complete power plant from the reactor neutronics to the electricity generation and distribution. All primary, secondary, and auxiliary systems are modeled including all control logic in order to provide the most accurate representation of actual nuclear power plant (NPP) operation, and the simulator results have been rigorously verified by an actual NPP operating at approximately 1 GWe. This report is a continuation of worked performed in previous years, and supplemental information from previous reports is included in the appendix for reference.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

United States–Brazil Joint Study: A Preliminary Assessment of Opportunities and Challenges for Small Modular Reactors in Brazil

This report supplies the technical information to enable a broader understanding of different technology choices for nuclear reactors in Brazil. The assessment characteristics of this process include reactor technology, economic costs for deployment, and regulatory analyses of small modular reactors, advanced reactors, and microreactors. The analysis documented in this report provides economic and technological information to assist Brazil, whose official indicative scenarios show an increase in operating nuclear power from 2 Gigawatt in 2022 to 8–10 Gigawatt by 2050, in policy making and long-term energy planning. To protect company proprietary information, the information in this report was limited to publicly available information and therefore can only inform preliminary reactor technology assessments. This report does not recommend or endorse any specific reactor developer or technology.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Anaerobic sequencing batch reactor in pilot scale for treatment of tofu industry wastewater

The small industry of tofu production process releases the waste water without being processed first, and the wastewater is directly discharged into water. In this study, Anaerobic Sequencing Batch Reactor in Pilot Scale for Treatment of Tofu Industry was developed through an anaerobic process to produce biogas as one kind of environmentally friendly renewable energy which can be developed into the countryside. The purpose of this study was to examine the fundamental characteristics of organic matter elimination of industrial wastewater with small tofu effective method and utilize anaerobic active sludge with Anaerobic Sequencing Bath Reactor (ASBR) to get rural biogas as an energy source. The first factor is the amount of the active sludge concentration which functions as the decomposers of organic matter and controlling selectivity allowance to degrade organic matter. The second factor is that HRT is the average period required substrate to react with the bacteria in the Anaerobic Sequencing Bath Reactor (ASBR).The results of processing the waste of tofu production industry using ASBR reactor with active sludge additions as starter generates cumulative volume of 5814.4 mL at HRT 5 days so that in this study it is obtained the conversion 0.16 L of CH{sub 4}/g COD and produce biogas containing of CH{sub 4}: 81.23% and CO{sub 2}: 16.12%. The wastewater treatment of tofu production using ASBR reactor is able to produce renewable energy that has economic value as well as environmentally friendly by nature.

09 BIOMASS FUELS↗
DOE OSTI2012

Prompt Neutron Lifetime for the NBSR Reactor

In preparation for the proposed conversion of the National Institute of Standards and Technology (NIST) research reactor (NBSR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel, certain point kinetics parameters must be calculated. We report here values of the prompt neutron lifetime that have been calculated using three independent methods. All three sets of calculations demonstrate that the prompt neutron lifetime is shorter for the LEU fuel when compared to the HEU fuel and longer for the equilibrium end-of-cycle (EOC) condition when compared to the equilibrium startup (SU) condition for both the HEU and LEU fuels.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

Non-destructive assay of mechanical components using gamma-rays and thermal neutrons

This work presents the results obtained in the inspection of several mechanical components through neutron and gamma-ray transmission radiography. The 4.46 Multiplication-Sign 10{sup 5} n.cm{sup -2}.s{sup -1} thermal neutron flux available at the main port of the Argonauta research reactor in Instituto de Engenharia Nuclear has been used as source for the neutron radiographic imaging. The 412 keV {gamma}-ray emitted by {sup 198}Au, also produced in that reactor, has been used as interrogation agent for the gamma radiography. Imaging Plates - IP specifically designed to operate with thermal neutrons or with X-rays have been employed as detectors and storage devices for each of these radiations.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Challenges and Opportunities in Staffing and Knowledge Transfer at Research Reactors

University research reactors have been a cornerstone of nuclear engineering research and education since the first reactor was deployed at North Carolina State University in the 1950s. The population of university reactors grew to a high of almost 80 in 1970 but has dropped to 24 operating today. The US Department of Energy – Office of Nuclear Energy supports the remaining reactors through fuel and infrastructure funding. The Nuclear Science User Facilities (NSUF) collaborated with the National Organization of Test, Research, and Training Reactors (TRTR) to study the needs of the university research reactor community. Staffing and knowledge transfer were identified as critical areas. The panel will discuss challenges and opportunities in these areas including how COVID-19 has affected their facilities.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

ADAPTATION OF CRACK GROWTH DETECTION TECHNIQUES TO US MATERIAL TEST REACTORS

A key component in evaluating the ability of Light Water Reactors to operate beyond 60 years is characterizing the degradation of materials exposed to radiation and various water chemistries. Of particular concern is the response of reactor materials to Irradiation Assisted Stress Corrosion Cracking (IASCC). Some test reactors outside the United States, such as the Halden Boiling Water Reactor (HBWR), have developed techniques to measure crack growth propagation during irradiation. The basic approach is to use a custom-designed compact loading mechanism to stress the specimen during irradiation, while the crack in the specimen is monitored in-situ using the Direct Current Potential Drop (DCPD) method. In 2012 the US Department of Energy commissioned the Idaho National Laboratory and the MIT Nuclear Reactor Laboratory (MIT NRL) to take the basic concepts developed at the HBWR and adapt them to a test rig capable of conducting in-pile IASCC tests in US Material Test Reactors. The first two and half years of the project consisted of designing and testing the loader mechanism, testing individual components of the in-pile rig and electronic support equipment, and autoclave testing of the rig design prior to insertion in the MIT Reactor. The load was applied to the specimen by means of a scissor like mechanism, actuated by a miniature metal bellows driven by pneumatic pressure and sized to fit within the small in-core irradiation volume. In addition to the loader design, technical challenges included developing robust connections to the specimen for the applied current and voltage measurements, appropriate ceramic insulating materials that can endure the LWR environment, dealing with the high electromagnetic noise environment of a reactor core at full power, and accommodating material property changes in the specimen, due primarily to fast neutron damage, which change the specimen resistance without additional crack growth. The project culminated with an in-pile demonstration at the MIT Reactor. The test rig and associated support equipment were used to apply loads to a representative Compact Tensile specimen during one MITR operating cycle, while measuring crack growth using the DCPD method. Although the test period was short (approximately 70 days), and the accumulated neutron dose relatively small, successful operation of the test rig was demonstrated. The specimen was cycled more than 8000 times (more than would be typical for a long term IASCC test), which was sufficient to propagate a crack of over 2 mm.

Crack Growth↗
DOE OSTI2015

Evaluation of an unexpected reflector temperature effect for water-isolated array configurations - 14500

It has been conventional wisdom in the criticality safety discipline that for spent fuel systems like water pit storage and transportation, modeling water at its peak density will maximize reactivity. Thus, modeling water at 3.96 deg. C will produce a conservatively bounding k{sub eff} for that configuration. This assumption has been found to be incorrect for certain high-leakage water-isolated array configurations. In fact, temperatures up to the boiling point of water can in fact be more reactive, and the most reactive temperature is highly dependent upon the specific fuel configuration being analyzed. In light of the accident at Fukushima-Daiichi - where the spent fuel pool temperature increased - knowledge of this unexpected behavior is essential to ensuring that criticality safety evaluations are conservatively bounding. To illustrate the reflector temperature effect, MC21 calculations are provided for a sensitivity study based on LEU-COMP-THERM-044, which models the IPEN/MB-01 reactor at various temperatures between 3.9 deg. C and 93.3 deg. C. (authors)

42 ENGINEERING↗
NASA NTRS1973

Earth storable bimodal engine, phase 1

An in-depth study of an Earth Storable Bimodal (ESB) Engine using earth storable propellants N2O/N2H4 and operating in either a monopropellant or bipropellant mode was conducted. Detailed studies were completed for both a hot-gas, regeneratively cooled thrust chamber and a ducted hot-gas, film cooled thrust chamber. Hydrazine decomposition products were used for cooling in either configuration. The various arrangements and configurations of hydrazine reactors, secondary injectors, chambers and gimbal methods were considered. The two basic materials selected for the major components were columbium alloys and L-605. The secondary injector types considered were previously demonstrated by JPL and consisted of a liquid-on-gas triplet, a liquid-on-gas doublet, and a liquid-on-gas coaxial injector. Various design tradeoffs were made with different reactor types located at: the secondary injector station, the thrust chamber throat, and the nozzle/extension interface. Associated thermal, structural, and mass analyses were completed.

Source record↗
DOE OSTI2022

A revisitation of the ENEA TAPIRO Fast Source Reactor neutronic capabilities

This work is devoted to the analysis of the neutronic characterization of the ENEA-Casaccia center RSV TAPIRO fast neutrons source reactor performed during the 80's in the frame of a joint ENEA-SCK-CEN Mol collaboration. The Belgian scientist Albert Fabry, who was available to investigate a fast research reactor to perform a full characterization of the system neutron flux, led the TAPIRO experimental campaign. In particular, the paper is focused on the comparison between some 'recommended' Fabry experimental reaction rates traverses and nowadays MCNP Monte Carlo and ERANOS deterministic calculation results. It appears that the quality of the results depends on the penetration depth into the copper reflector. The influence of copper nuclear data are more evident in the ERANOS results. Massimo Salvatores has provided a great contribution to the interpretation of the TAPIRO neutronics: this humble work is dedicated to him.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

RAVEN Template for Dynamic Representativity Analysis of the High Temperature Test Facility

These slides present a walkthrough of the template that has been developed for using RAVEN to perform representativity analysis using models of the High Temperature Test Facility and the General Atomics Modular High Temperature Gas-cooled Reactor. The presentation provides participants in the HTTF benchmark with a walkthrough on how to use RAVEN for their sensitivity analysis and how to read results from the MHTGR-350 to perform representativity

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗