AMS Joint/Bilateral Surveys
A historical presentation depicting the International Joint Aerial Surveys AMS has co-created and collaborative efforts that were implemented as result of those efforts.
61 RADIATION PROTECTION AND DOSIMETRY↗
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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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Explicit Dynamic Impact Analysis of the Building 3525 Packages
This report describes the methodology used for the safety basis analysis of the Building (Bldg.) 3525 casks and presents preliminary results for a selected geometry. The specific problem analyzed is a free fall load case onto an unyielding target at various orientations. The various orientations were analyzed to ensure that the packages were subjected to the orientation that causes the most damage. The analysis approach, material modeling, finite element modeling, and results for the selected case are presented, along with a brief discussion of conservatism and uncertainties. The results for the analyzed load cases show that through selected administrative controls, no breach of the container or loss of shielding will occur in the package.
42 ENGINEERING↗
Determining Aerosol Deposition on Detector Faces and Internal Surfaces in Continuous Air Monitors (In Addition to Measurements of Their Filter Collection Efficiencies)
One function of Continuous Air Monitor (CAM) samplers at LANL (Los Alamos National Laboratory) is to alert workers to the presence of airborne alpha-emitting (e.g.) plutonium particles.
61 RADIATION PROTECTION AND DOSIMETRY↗
1FRAME - 1F Fuel Retrieval and Monitoring Experiments [Poster]
The 1FRAME (1F Fuel Retrieval and Monitoring Experiments) project includes collaborative research and development in the area of neutron detection, analysis, and simulations for fuel debris removal at Fukushima Daiichi nuclear power station. Technical advances are needed in all three focus areas to provide technical recommendations on a course of action for fuel debris removal. This work is part of collaborative research and development efforts between the US, Japan, and France.
Fuel Debris Removal↗
Improved PARE (Pressurized Airborne Release Equipment) for ARF (Airborne Release Fraction) and RF (Respirable Fraction) Testing
An improved version of the PARE (Pressurized Airborne Release Equipment) system (Sutter 1983a and Sutter 1983b and Ballinger et al 1987) has been developed for powder venting testing. This effort seeks to improve the accuracy of risk estimation for the venting of pressurized powders and the venting of gases through pressurized powders.
61 RADIATION PROTECTION AND DOSIMETRY↗
Advanced Test Reactor Neutron Dosimetry Report - 173B (Rev. 1)
Results of neutron dosimetry measurements in ATR for cycle 173B-1&2 are presented. Cobalt and nickel wires were installed in numerous irradiation locations for the duration of the cycle. The specific activity of these wire segments can be used in conjunction with the presented cycle history information to determine the thermal- and fast-neutron fluence rates for each irradiation position during the cycle using the referenced standard test methods.
173B↗
Microreactor Incident-Free Transportation Radiation Dose Assessment
This report, prepared by Pacific Northwest National Laboratory (PNNL) for the U.S. Nuclear Regulatory Commission, provides an estimate of the radiation dose related to the transportation of microreactors following the regulatory limits established in Title 10 of the Code of Federal Regulations (10 CFR) 71.47(b). The transportation of irradiated microreactors poses unique challenges. Microreactors may be challenged to meet the current 10 mrem/hr limit at 2 meters from the vehicle in 10 CFR 71.47(b). The need to modify packaging to be within practical weight and size limits for shipping may limit the amount of shielding that can be applied for a microreactor transportation package design. Operational strategies that would necessitate moving a reactor within a short time after its operating life may result in higher radiation dose rates associated with lower cooling times. This study estimates the radiation dose to maximally exposed individuals (MEIs) under the current regulatory radiation level limit of 10 mrem/hr at 2 meters from the vehicle as well as alternative limits of 10, 50, 100, and 200 mrem/hr at 2 meters from the vehicle. This analysis expands on prior incident free transportation radiation dose assessments conducted for spent nuclear fuel (SNF) and radioactive waste transportation. The key factors analyzed include the microreactor fuel type and burnup, shipment cooling periods, shielding designs integrated into transport packaging, and exposure scenarios for MEIs for highway and rail transport. This report considers additional distances that may be associated with compensatory measures to ensure safe transport conditions under varying radiation dose limits, including enhanced routing measures, escort protocols, vehicle inspections, and operational restrictions.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
ACRRF High-Bay Dose Calculations using MCNP (Part A)
Analytical tools and models have been developed as a starting point for directly assessing dose in the Annular Core Research Reactor Facility (ACRRF) due to reactor operation. Key results include peak dose along the Central Cavity (CC) centerline (beamline) at the cavity level, dose throughout the High-Bay (HB), and dose on the facility roof for partially-shielded reactor operation where the 4” insert is removed from the CC Shield Plug (SP). Model results in the beamline are benchmarked against measured doses from passive dosimetry evaluations. Personnel total (neutron and gamma) dose in the ACRRF HB is calculated using Monte Carlo N-Particle (MCNP). Various CC and SP configurations are analyzed, including unshielded (no SP) and partially shielded (SP installed but 4” insert removed). Novel application of Variance Reduction (VR) techniques, namely the Surface Source Write (SSW) and Surface Source Read (SSR) capabilities in MCNP, enable impressive resolution (in a Monte Carlo modeling sense) of dose throughout much the facility. The VR techniques reduce stochastic error for challenging tallies, with more advanced techniques explored in the companion to this report (Part B) [1]. Supplementary studies (including a verification analysis) and pedagogic evaluations in Part B involve neutron spectra, angular distributions, and the dose impact of facility characteristics. With the SP 4” insert removed and the Lead-Boron (44”) Bucket (LB–44) in the reactor cavity, Total Effective Dose (TED) within the CC beamline is ≈140 rem per 300 MJ of reactor yield (or 3900 rem per hour at 100% Steady-State (SS) power). With no SP (unshielded) and a Free-Field (FF) cavity, TED within the beamline is ≈610 rem per 300 MJ (or 17000 rem per hour at 100% SS power). Due to the predicted collimation of radiation by the reactor pool (and partial SP, if present), beamline dose is much greater than the scattered radiation field surrounding the cavity and reactor tank. Comparisons are made to beamline dosimetry measurements to validate the model. Model predictions agree reasonably well (⪅10%) with measured quantities of neutron fluence, gamma fluence, and spectral metrics. Away from the beamline, comparisons made to previous dose measurements in the HB agreement within an order of magnitude.
61 RADIATION PROTECTION AND DOSIMETRY↗
ACRRF High-Bay Dose Calculations using MCNP (Part B)
ACRR radiation outputs through vertical cavities have been documented in two reports. In Part B, the maximum dose from the unshielded central cavity, FREC-II, and NRS is calculated to inform the safety basis. The maximum dose is ≈1430 mrem per 300 MJ, or ≈40,930 rem/hr for full-power operation. A verification study completes the V&V of the modeling. Supplementary studies of variance-reduction techniques, model sensitivities, and aircraft dose above the ACRRF are included.
61 RADIATION PROTECTION AND DOSIMETRY↗
User’s Manual for RESRAD-RDD&IND Code Version 2: Vol. 1—Methodology and Models Used in RESRAD-RDD&IND Code
RESRAD-RDD&IND is part of the RESRAD family of codes that Argonne National Laboratory developed for the U.S. Department of Energy (DOE). An earlier version, RESRAD-RDD published in 2009, dealt only with radiological dispersal device (RDD) incidents (DOE 2009). This new version, RESRAD-RDD&IND, is designed to evaluate both RDD incidents and improvised nuclear device (IND) incidents. This report is Volume 1 of the RESRAD-RDD&IND User’s Manual that documents the methodology, models, and radionuclide-specific data used in RESRAD-RDD&IND code Version 2.0. Volume 2 of the RESRAD-RDD&IND User’s Manual is called the User’s Guide for RESRAD-RDD&IND Code . It describes how to use RESRAD-RDD&IND code Version 2.0 and includes screen shots and parameter information.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Worst Case Sabotage Consequence Modeling for Advanced Reactors
Under proposed rulemaking by the United States Nuclear Regulatory Commission, advanced reactor licensees may qualify for a reduction in onsite responders if the consequences from a sabotage attack do not exceed 25 rem at the reactor site boundary 2 hours after a release occurs. This work presents a preliminary assessment of the feasibility of meeting the 25 rem at the site boundary criteria. Using advanced reactor radionuclide inventories radiological release estimates created by collaborators at Oak Ridge National Laboratory, the MACCS consequence analysis code was used to investigate the distance at the dose from hypothetical sabotage events reaches 25 rem for different advanced reactor technologies. It was found that the distance to 25 rem depended heavily on sampled weather conditions and sabotage release fractions, and further refinement of release fractions and MACCS modeling parameters is recommended.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Advanced Test Reactor Neutron Dosimetry Report - Cycle 175B-1
Results of neutron dosimetry measurements in ATR for cycle 175B-1 are presented. A single set of cobalt and nickel wires were installed in numerous irradiation locations for the duration of the cycle. The specific activity of these wire segments can be used in conjunction with the presented cycle history information to determine the thermal- and fast-neutron fluence rates for each irradiation position during the cycle using the referenced standard test methods.
175B↗
Redesign of Nuclear Material Storage Container with Additive Manufacturing
The goal of this project is to reduce worker dose and improve criticality safety by improving nuclear material storage containers. The current method of medium to long term (up to 15 years) storage of small amounts of nuclear material is to store it in a stainless-steel flat tube with a copper crush gasket. These containers use 10 bolts to fasten the lid, and the crush gasket must be replaced after every use. Stainless-steel acts as a gamma shield, but there is not any material to account for neutron dose. A redesign of this container should prioritize reducing radiation doses to handlers of the container by reducing the amount of time handlers are exposed to the container and introducing material with effective neutron and gamma shielding mechanics.
42 ENGINEERING↗
Application & Testing of the Hidex ULLA at LANL
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07 ISOTOPE AND RADIATION SOURCES↗
Augmentation of Radiological Control Technicians (RCTs) for Truck Surveys
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42 ENGINEERING↗
Multi-Agency Radiation Survey and Assessment of Materials and Equipment (MARSAME) Radiological Release Report for Technical Area 33 Decontamination and Demolition Project
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54 ENVIRONMENTAL SCIENCES↗
Nuclear Safety [Vol. 37, No. 1, January-March 1996]
Nuclear Safety is a journal that covers significant issues in the field of nuclear safety. Its primary scope is safety in the design, construction, operation, and decommissioning of nuclear power reactors worldwide and the research and analysis activities that promote this goal, but it also encompasses the safety aspects of the entire nuclear fuel cycle, including fuel fabrication, spent-fuel processing and handling, and nuclear waste disposal, the handling of fissionable materials and radioisotopes, and the environmental effects of all these activities. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 1 A Decision Support System for Maintenance Management of a Boiling-Water Reactor Power Plant, J. H. Shen, A. Ray, and S. Levine; ACCIDENT ANALYSIS: 12 On Prediction of the Ignition Potential of Uranium Metal and Hydride, M. Epstein, W. Luangdilok, M. G. Plys, and H. K. Fauske; 26 An Overview of the Primary Parameters and Methods for Determining Condensation Heat Transfer to Containment Structures, J. Green and K. Almenas; DESIGN FEATURES: 49 Modem Tornado Design of Nuclear and Other Potentially Hazardous Facilities, J. D. Stevenson and Y. Zaho; U.S. NUCLEAR REGULATORY COMMISSION INFORMATION AND ANALYSES: 73 1994 Accident Sequence Precursor Program Results, R. J. Belles, J. W. Cletcher, D. A. Coplnger, B. W. Dolan, J. W. Minarick, and P. D. O'Reilly; ANNOUNCEMENTS: 93 American Institute of Chemical Engineers (AlChE) Spring 1997 National Meeting; 94 European Safety and Reliability Association International Conference on Safety and Reliability ESREL ’97; 95 Criticality Safety Challenges in the Next Decade; 96 21st International Symposium on the Scientific Basis for Nuclear Waste Management; 84 The Authors; 88 Letter to the Editor; 90 Indexes to Nuclear Safety, Volume 36.
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Nuclear Safety [Vol. 37, No. 2, April-June 1996]
Nuclear Safety is a journal that covers significant issues in the field of nuclear safety. Its primary scope is safety in the design, construction, operation, and decommissioning of nuclear power reactors worldwide and the research and analysis activities that promote this goal, but it also encompasses the safety aspects of the entire nuclear fuel cycle, including fuel fabrication, spent-fuel processing and handling, and nuclear waste disposal, the handling of fissionable materials and radioisotopes, and the environmental effects of all these activities. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 97 The Nuclear Community and the Public: Cognitive and Cultural Influences on Thinking About Nuclear Risk, M. A. Meyer; 109 Twenty-Third Water Reactor Safety Information Meeting, D. A. Copinger; ACCIDENT ANALYSIS: 126 Analysis of a PWR LBLOCA Without SCRAM, Trevor N. Tyler, Rafael Macian-Juan and John H. Mahaffy; DESIGN FEATURES: 139 Vulnerability of Multiple-Barrier Systems, N. C. Lind; ENVIRONMENTAL EFFECTS: 149 A Study of Wet Catalytic Oxidation of Radioactive Spent Ion Exchange Resin by Hydrogen Peroxide, Xingchao Jian, Tianbao Wu, and Guichun Yun; 157 A Comparison Study and Resolution of Differences Between Emergency Response and Safety Analysis Codes Used at the Savannah River Site, A. A. Simpkins; OPERATING EXPERIENCES: 164 Reactor Shutdown Experience, Compiled by J. W. Cletcher; RECENT DEVELOPMENTS: 167 Reports, Standards, and Safety Guides, D. S. Queener; 172 Proposed Rule Changes as of Dec. 31,1995; ANNOUNCEMENTS: 178 American Nuclear Society 1997 Annual Meeting; 178 American Nuclear Society Nuclear Criticality and Safety Division Topical Meeting; 176 The Authors.
05 NUCLEAR FUELS↗