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At least 1,225 records · Page 68
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

SCALE Code System

The SCALE Code System is a widely-used modeling and simulation suite for nuclear safety analysis and design that is developed, maintained, tested, and managed by the Reactor and Nuclear Systems Division (RNSD) of Oak Ridge National Laboratory (ORNL). SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor and lattice physics, radiation shielding, spent fuel and radioactive source term characterization, and sensitivity and uncertainty analysis. Since 1980, regulators, licensees, and research institutions around the world have used SCALE for safety analysis and design. SCALE provides an integrated framework with dozens of computational modules including three deterministic and three Monte Carlo radiation transport solvers that are selected based on the desired solution strategy. SCALE includes current nuclear data libraries and problem-dependent processing tools for continuous-energy (CE) and multigroup (MG) neutronics and coupled neutron-gamma calculations, as well as activation, depletion, and decay calculations. SCALE includes unique capabilities for automated variance reduction for shielding calculations, as well as sensitivity and uncertainty analysis. SCALE’s graphical user interfaces assist with accurate system modeling, visualization of nuclear data, and convenient access to desired results.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2020

SCALE Code System

The SCALE Code System is a widely used modeling and simulation suite for nuclear safety analysis and design that is developed, maintained, tested, and managed by the Reactor and Nuclear Systems Division (RNSD) of Oak Ridge National Laboratory (ORNL). SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor and lattice physics, radiation shielding, spent fuel and radioactive source term characterization, and sensitivity and uncertainty analysis. Since 1980, regulators, licensees, and research institutions around the world have used SCALE for safety analysis and design. SCALE provides an integrated framework with dozens of computational modules, including three deterministic and three Monte Carlo radiation transport solvers that are selected based on the desired solution strategy. SCALE includes current nuclear data libraries and problem-dependent processing tools for continuous-energy (CE) and multigroup (MG) neutronics and coupled neutron-gamma calculations, as well as activation, depletion, and decay calculations. SCALE includes unique capabilities for automated variance reduction for shielding calculations, as well as sensitivity and uncertainty analysis. SCALE’s graphical user interfaces assist with accurate system modeling, visualization of nuclear data, and convenient access to desired results.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Revised decay properties of the key 93-keV resonance in the Mg 25 ( p , γ ) reaction and its influence on the MgAl cycle in astrophysical environments

The γ-decay properties of an excited state in 26 Al at 6398.3(8) keV have been reexamined using the 11 B+ 16 O fusion-evaporation reaction. This level represents a key 93.1(8)-keV resonance in the 25 Mg+p system and its relative branching to the 26 Al ground state, f 0 , has been determined to be 0.76 ± 0.03 (stat.) ± 0.10 (syst.). Furthermore, this is a significantly higher value than the most recent evaluation and implies a considerable increase in the production of cosmic γ rays from 26 Al radioactivity.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2026

Nuclear Radii of Proton-Unbound Systems

Nuclear radius is a fundamental structural observable that informs many properties of atomic nuclei and nuclear matter. Experimental studies of radii in drip line nuclei are in the forefront of research with radioactive ion beams. Of particular interest are charge radii of proton-unbound nuclei that will soon be approached in laser spectroscopy. In this Letter, using the complex-energy approach and direct time propagation, we investigate the radius of the proton resonance whose size is ill defined in the standard stationary quantum-mechanical description. An early-time plateau is identified during which the radius of the Gamow resonance coincides with the real-energy radius accessible experimentally. We demonstrate a nonmonotonic dependence of the complex radius on decay energy and a local increase of the charge radius across the threshold (a halolike enhancement).

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Permanent Criticality Termination Processes in Disposed DPCs

This report aims at answering what, how, and when spent nuclear fuel (SNF) or dual-purpose canister (DPC) characteristics could be impacted by disposal events and processes, including decay, corrosion, dissolution, and criticality, such that the potential for criticality initiation or continuation in disposed DPCs becomes permanently significantly diminished. This report uses the term "permanent termination of criticality to denote the significant diminishment of criticality potential, not absolute prevention. The occurrence of disposal processes and events is a direct function of disposal time. For fundamental processes (e.g., decay), time is absolute; however, for other processes (e.g., corrosion), time is relative because it is driven by a combination of DPC characteristics (e.g., fuel conditions, basket composition), geologic parameters (e.g., infiltration rate), engineered barrier design, and other processes and events that impact in-package chemistry.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Accelerating Surface Radiation Content to Investigate the Impact of Radon Progeny on Superconducting Qubits

Ionizing radiation in the form of $α$, $β$, $γ$, and additional high-energy particles can induce decoherence via phonon and quasiparticle poisoning in superconducting qubits. Recent studies have explored this effect using cosmic rays or controlled radioactive sources held in the proximity of a qubit package, and have concluded that reductions in such ``external'' environmental radiation may benefit stable operation of qubit devices. However, the effect of long-lived, unstable daughters of $^{222}$Rn that ``plate out'' directly on device and packaging surfaces has not been as extensively explored. This plate-out process, well-known to the dark matter direct detection field, occurs throughout the fabrication and testing lifecycle of a device and (separately) its packaging, and produces a local source of $α$-decays which can remain active for decades. As this scales with chip area, understanding and managing this source of ionizing radiation is relevant for successfully scaling quantum computing architectures to larger numbers of qubits in a radiation-robust way. We present a setup capable of accelerating and enhancing radon daughter plateout by a factor of $7\times10^4$ over ambient, in order to study, \textit{in situ}, the impact of these events on superconducting qubits. We also provide outlook on the potential impact of this source of ionizing radiation on current and future qubit arrays.

Poudel, Sagar S. [South Dakota Sch. Mines Tech.]↗
DOE OSTI2014

Final Technical Report for Award DE-FG02-98ER41080

The prime motivation of the project at McMaster University was to carry out the critical evaluation and compilation of Nuclear Structure and Decay data, and of nuclear astrophysics data with continued participation in the United States Nuclear Data Program (US-NDP). A large body of evaluated and compiled structure data were supplied for databases such as ENSDF, XUNDL, NSR, etc. residing on webpage of National Nuclear Data Center of the Brookhaven National Laboratory, Upton, New York, USA. Thermonuclear reaction rates of importance to stellar explosions, such as novae, x-ray bursts and supernovae, were evaluated as well. This effort was closely coupled to our ongoing experimental effort, which took advantage of radioactive ion beam and stable beam facilities worldwide to study these key reaction rates. This report contains brief descriptions of the various activities together with references to all the publications in peer-reviewed journals which were the result of work carried out with the award DE-FG02-98-ER41080, during 1998-2013.

Chen, Alan↗
DOE OSTI2020

In-situ Detection of Glovebox Glove Degradation Prior to Glove Failure; Effects on Safety, Waste, and Operational Costs - 20402

Glovebox glove pressure decay leak testing and analysis of the data can greatly improve glovebox safety, minimize Transuranic (TRU) glove waste, and minimize operational cost. It is well known, the weakest point of containment on a glovebox is the glove. Mitigating unplanned openings in gloves is critical to minimizing operational and safety costs in glovebox operations. Mitigating unplanned glove openings due to glove failures will be discussed in this paper. The lack of an engineered solution to determine a gloves operational life in nuclear operations, has lead to use of theoretical and manual subjective methods to determine the life and safety of a glovebox glove. Whether the glove is inspected visually prior to use, or gloves are changed regularly to theoretically avoid failure, both the operators PPE and the room are being exposed to contamination, or un-necessary amounts of TRU glove waste are being generated. Millions of dollars yearly in operational costs are spent on glove failures due to loss of production, incident analysis, regulatory audits, clean up, and paperwork. In a nuclear application, a glove failure could contaminate an operator and laboratory, shut down the operation for weeks, potentially create a regulatory audit, and potentially create a media frenzy; the damages to the organization can be astronomical. Elimination of glove failures is possible by tracking glove material degradation and setting limits to allow glove change when it becomes necessary, prior to glove failure. Analytical pressure decay leak testing allows the detection of material degradation, and in turn the ability to limit operational glove failures. A German based company called MK Versuchsanlagen, e.K., has developed an advanced Glove Integrity Testing System. The system, with its use of proprietary RFID and special software technology, is unique in its capabilities. The system is capable to perform highly accurate regular glove testing on any size glovebox line in minimal time. The system records and can analyze a tremendous amount of system data that in turn can determine, for example, a safe glove change time prior to a glove material failure. This paper will show how an advanced Glove Integrity Testing System can detect glove material degradation and help determine a safe change point prior to glove failure. Pressure decay curves of new gloves and recorded effects of accelerated aging on the glove over time will be shown. The use of this technology in nuclear facilities can greatly improve glovebox operational safety, minimize glovebox glove TRU waste, and save organizations tremendous costs in downtime, clean up, documentation, and potential regulatory review. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Effects of Mineral Impurities and Heat on Uranium(VI) Sorption onto Bentonite (Final Report)

Most nuclear waste disposal options include compacted bentonite, consisting of montmorillonite clay, as part of a barrier system to minimize contaminant mobility. Uranium (U) is the primary element in spent nuclear fuel, and from an environmental perspective, a potential contaminant of water resources. Furthermore, uranium may control nuclear fuel degradation rates and the consequent release of other radioactive contaminants based on its diffusive transport away from waste canisters. Uranium sorption onto clay and other mineral surfaces is expected to limit U(VI) mobility in these systems. However, at this point a prediction of U(VI) sorption and transport behavior in performance assessment (PA) models is complicated by a series of factors, such as: (1) the presence of various U(VI) solution species with different charges and sorption characteristics; (2) the complex microstructure of montmorillonite clay resulting in two types of clay porosities and multiple sorption sites; (3) the largely unknown effects of bentonite mineral impurities on pore water chemistry, U(VI) sorption and diffusion behavior; and (4) the potential impacts of heat, generated by the decay of spent fuel, on mineralogical and microstructural transformations, and any subsequent effects on radionuclide sorption and mobility. In this project, we investigated the effects of calcite impurities on U(VI) sorption and diffusion onto montmorillonite before and after mineral exposure to heat. Overall, our main goals were to: 1. Gain a mechanistic understanding of the effects of calcite impurities on U(VI) sorption and diffusion in montmorillonite/bentonite at variable chemical conditions, and before and after mineral exposure to heat; and to 2. Work towards the development of new surface complexation and diffusion models to predict U(VI) sorption and mobility in montmorillonite/bentonite while accounting for the impacts of calcite impurities and heat. In order to achieve these goals, we pursued the following research objectives and tasks: 1. Characterize the effects of calcite impurities on U(VI) sorption onto Na-montmorillonite, and evaluate if and how these effects change due to solids’ exposure to heat (Task 1). 2. Characterize U(VI) surface speciation in solid samples from Task 1, with a particular focus on U(VI) precipitates and co-precipitates (Task 2). 3. Test the relevance of observed changes in U(VI) sorption characteristics for U(VI) diffusive transport behavior in bentonite barriers (Task 3). 4. Evaluate the exclusion of anionic U(VI) solution complexes from clay nanopores (Task 4). 5. Building on Tasks 1-4, start the development of new surface complexation and reactive transport models that can simulate and predict U(VI) sorption and diffusion in complex bentonite systems (Task 5). A detailed description of all tasks, and their specific results and conclusions is provided in this report.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Off-Gas Generation Rate during Chemical Cleaning Operations at the Savannah River Site - 12499

The enhanced chemical cleaning process (ECC) is being developed at the Savannah River Site (SRS) to remove the residual radioactive sludge heel that remains in a liquid waste storage tank. Oxalic acid is the chemical agent utilized for this purpose. However, the acid also corrodes the carbon steel tank wall and cooling coils. If the oxalic acid has little interaction with the sludge, hydrogen gas could conceivably evolve at cathodic areas due to the corrosion of the carbon steel. Scenarios where hydrogen evolution could occur during ECC include the initial filling of the tank prior to agitation and near the end of the process when there is little or no sludge present. The purpose of this activity was to provide a bounding estimate for the hydrogen generation rate during the ECC process. Sealed vessel coupon tests were performed to estimate the hydrogen generation rate due to corrosion of carbon steel by oxalic acid. These tests determined the maximum instantaneous hydrogen generation rate, the rate at which the generation rate decays, and the total hydrogen generated. The tests were performed with polished ASTM A285 Grade C carbon steel coupons. This steel is representative of the Type I and II waste tanks at SRS. Bounding conditions were determined for the solution environment. The oxalic acid concentration was 2.5 wt.% and the test temperature was 75 deg. C. The test solution was agitated and contained no sludge simulant. Duplicate tests were performed and showed excellent reproducibility for the hydrogen generation rate and total hydrogen generated. The results showed that the hydrogen generation rate was initially high, but decayed rapidly within a couple of days. A statistical model was developed to predict the instantaneous hydrogen generation rate as a function of exposure time by combining both sets of data. An upper bound on the maximum hydrogen generation rate was determined from the upper 95% confidence limit. The upper bound limit on the maximum instantaneous generation rate at 5 hours was 6.1 x 10{sup -5} m{sup 3}/m{sup 2}/minute. After two and five days the upper bound limit decayed to 7.9 x 10{sup -6} and 1.3 x 10{sup -6} m{sup 3}/m{sup 2}/minute, respectively. The total volume of hydrogen gas generated during the test was calculated from the model equation. An upper bound on the total gas generated was determined from the upper 95% confidence limit. The upper bound limit on the total hydrogen generated during the 163 hour test was 0.101 m{sup 3}/m{sup 2}. Corrosion rates were determined from the coupon tests and also calculated from the measured hydrogen generation rates. Excellent agreement was achieved between the time averaged corrosion rate calculated from the hydrogen generation rates and the corrosion rates determined from the coupon tests. The corrosion rates were on the order of 0.45 mmpy. Good agreement was also observed between the maximum instantaneous corrosion rate as calculated from the hydrogen generation rate and the corrosion rate determined by previous electrochemical tests. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

2022 Report - SRNL Aging and Lifetimes program tritium aging studies on structural alloys

This report documents work performed in fiscal year 2022 at SRNL in support of the Aging and Lifetimes program. This work is an enduring collaboration between SRNL and SNL to study tritium embrittlement of structural metals used in Gas Transfer System reservoirs. The measured data inform component lifetime assessments and predictions. Test coupons are thermally pre-charged with tritium and allowed to age at cryogenic temperatures to freeze out tritium diffusion and minimize off-gassing while allowing for decay helium levels to build. Coupons are removed and tested at planned intervals to measure how mechanical properties degrade as helium levels increase. This report summarizes test results from coupons which are in various stages of aging as they were tritium pre-charged in previous years, as well as documenting one new charging run.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Implications of Plutonium isotopic separation on closed fuel cycles and repository design

Advances in laser enrichment may enable relatively low-cost plutonium isotopic separation. This would have large impacts on LWR closed fuel cycles and waste management. If Pu-240 is removed before recycling plutonium as mixed oxide (MOX) fuel, it would dramatically reduce the buildup of higher plutonium isotopes, Americium, and Curium. Pu-240 is a fertile material and thus can be replaced by U-238. Eliminating the higher plutonium isotopes in MOX fuel increases the Doppler feedback, simplifies reactor control, and allows infinite recycle of MOX plutonium in LWRs. Eliminating fertile Pu-240 and Pu-242 reduces the plutonium content in MOX fuel and simplifies fabrication. Reducing production of Pu-241 reduces production of Am-241 - the primary heat generator in spent nuclear fuels after several decades. Reducing heat generating Am-241 would reduce repository cost and waste toxicity. Avoiding Am- 241 avoids its decay product Np-237, a nuclide that partly controls long-term oxidizing repository performance. Most of these benefits also apply to LWR plutonium recycled into fast reactors. There are benefits for plutonium isotopic separation in fast reactor fuel cycles (particularly removal of Pu-242) but the benefits are less. (author)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Intrinsic dosimetry of glass containers: A potential interrogation tool for nuclear forensics and waste management

Intrinsic dosimetry is the method of measuring total absorbed dose received by the walls of a container holding radioactive material. By considering the total absorbed dose received by a container in tandem with the physical characteristics of the radioactive material housed within that container, this method has the potential to provide law enforcement officials or waste management personnel with enhanced pathway information regarding the history of the container and its radioactive contents. We report the latest in a series of experiments designed to validate and demonstrate this newly developed tool. Thermoluminescence (TL) dosimetry was used to measure dose effects on the raw stock material of borosilicate container glass with time up to 60 days after gamma ray, x-ray, beta particle or ultraviolet irradiations at doses from 0.15 to 20 Gy. Two peaks were identified in the TL glow curve when irradiated with 60Co, a relatively unstable peak around 120°C and a more stable peak around 225°C. Signal strength of both peaks decayed with time. The minimum measurable dose using this technique was measured to be 0.15 Gy, which is roughly equivalent to a 24 hr irradiation at 1 cm from a 50 ng source of 60Co. After fading, this dose would still be observable approximately 1 year post-irradiation. Similar doses with beta particles and x-rays produced TL glow curves of lower intensity. The TL glow curves were separated into five peaks centered at approximately 120, 160, 225, 300, and 340°C. Differences in TL glow curve shape and intensity were observed for the glasses from different geographical origins. These differences can be explained by changes in the intensities of the five peaks. EPR results showed production of BOHC and E’ silica centers.

Clark, Richard A.↗
DOE OSTI2022

Air Exchange Rate Impact on Activity Equilibrium Factors and Inhalation Fractional Equilibrium Factors for Rn, Xe, Kr, Ar, Ne, and Their Progeny in Vapor Intrusion, Risk, and Dose Models

Exposure to the radioactive noble gasses, especially radon, is of high concern and poses a significant risk to humans in an indoor air environment as the second leading cause of lung cancer in the United States. To evaluate and minimize the risks posed by these gasses, it is important to understand their radiological and physical properties. The EPA’s Radon Vapor Intrusion Screening Level (RVISL) calculator calculates indoor air RVISLs based on target working levels (WLs), target excess lifetime cancer risk (ELCR), and annual dose limits for the actinon (Rn-219), thoron (Rn-220), and radon (Rn-222) decay series. The RVISLs are based on inhalation and submersion in gas cloud exposure routes for residential and commercial settings. The RVISLs are analogous to preliminary remediation goals (PRGs) and dose compliance concentrations (DCCs), where the isotope-specific values are in units of activity concentration (activity per unit volume). If the concentration of a parent isotope of radon or its progeny is found to exceed the RVISL, then further action to ensure cleanup of the contaminant may be necessary. In residential and commercial settings, the RVISLs will vary based on the air exchange rate present. The EPA’s Radionuclide PRG and DCC Calculators also assess the risk/dose from noble gases in the air due to household use of water like showering. In this study, a computational method in MATLAB was developed to determine the impact of the air exchange rate on the activity equilibrium factor (A eq ) and the inhalation fractional equilibrium factor (F eq ). Both factors are values that reflect the equilibrium concentrations of progeny to their parent in the air. These factors have a direct impact on the RVISL, PRG, and DCC calculations of WL, ELCR, and annual dose, respectively. This study builds on a previous report that only focused on actinon, thoron, and radon by revisiting the original A eq and F eq calculation methods, as well as including the values for the Rn-207, Rn-209, Rn-210, Rn-211, Rn-215, Rn-216, Rn-217, Rn-218, Rn-223, Ne-24, Ar-42, Ar-43, Ar-44, Kr-74, Kr-75, Kr-76, Kr-77, Kr-88, Kr-89, Xe-120, Xe-121, Xe-122, Xe-123, Xe-135m, and Xe-138 decay chains, which are not currently available in literature. The EPA’s RVISL calculator will be updated to include the new A eq and F eq values for the actinon, thoron, and radon decay chains, while the rest of the calculators will incorporate all the new A eq and F eq values as appropriate.

63 RADIATION, THERMAL, AND OTHER ENVIRON. POLLUTAN↗
DOE OSTI2015

Ra-226: An Inter-Method Comparison - 15447

Radium is an element which can be found naturally at low levels in the environment (water, rock, soil) as well as due to impact from man. It is known to cause cancer, primarily through ingestion and inhalation, and is considered one of the main radioactive environmental concerns. Sources of ingestion include food and drinking water, while sources of inhalation include emissions from the burning of coal or other fuels as well as dust generated from the uranium or phosphate mining industries. Radium is generally present in relatively low levels in the environment, but certain geographic formations containing elevated concentrations of radium exist throughout the world. In addition, there are sites which are considered contaminated by prior industrial activity. Radium's most prevalent isotopes are Ra-224, Ra-226, and Ra-228. Ra-226, a member of the U-238 decay series with a half-life of 1600 years, is the most common radium isotope. Several methods exist for analysis of this isotope, some targeting measurement of this radionuclide's alpha emission (or those of its decay progeny), others focused on the accompanying gamma-ray emissions. The choice of analytical method to utilize is dependent upon various factors. The purpose of this discussion is to identify various methods, highlighting strengths and weaknesses of each to assist in determining the most advantageous method to utilize, and to compare results obtained from each method. (authors)

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2021

Measurement of the ionization response of amorphous selenium with 122 keV γ rays

In this work, we performed a measurement of the ionization response of 200 μm-thick amorphous selenium (aSe) layers under drift electric fields of up to 50 V/μm. The aSe target was exposed to ionizing radiation from a 57 Co radioactive source and the ionization pulses were recorded with high resolution. Using the spectral line from the photoabsorption of 122 keV γ rays, we measured the charge yield in aSe and the line width as a function of drift electric field. From a detailed microphysics simulation of charge generation and recombination in aSe, we conclude that the strong dependence of recombination on the ionization track density provides the dominant contribution to the energy resolution in aSe. These results provide valuable input to estimate the sensitivity of a proposed next-generation search for the neutrinoless ββ decay of 82 Se that aims to employ imaging sensors with an active layer of aSe. We estimate the RMS line width of the integrated ionization signal from neutrinoless ββ decay events (of deposited energy 3 MeV) to be 2.0% for a drift field of 50 V/μm. The energy resolution can be improved to 1% by correcting for the charge yield as a function of ionization density along the imaged electron tracks.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

{sup 99}Mo Production at KIPT Neutron Source Facility

Argonne National Laboratory (ANL) and Kharkov Institute of Physics and Technology (KIPT) have cooperated on the construction of the Neutron Source Facility (NSF) at KIPT, Kharkov (Ukraine). The NSF facility consists of an electron accelerator driven system (ADS) generating 300 kW of thermal power for multiple applications, including medical isotope production. ANL performed extensive analyses including medical isotopes production, reactivity monitoring, operations and control, refueling, cold neutron source design, target design,and radiation shielding. The present study focuses on the use of the KIPT NSF for medical isotope production. Medical isotopes are radioactive isotopes with short half-lives used in diagnostic and therapeutic medical procedures. For example, brain cancer is difficult to treat by ordinary surgery and only brachytherapy with {sup 131}Cs, which has 9.7 days half-life, has shown effective results on the patients. {sup 99m}Tc is the most common medical isotope used in medical diagnoses because the wavelength of its gamma radiation (8.8 pm) is about the same as the conventional X-ray diagnostic devices. {sup 99m}Tc has 6 hours half-life, therefore it cannot be stored because it decays into {sup 99}Tc. The parent of {sup 99m}Tc is {sup 99}Mo, which has 66 hours half-life. Consequently, the delivery of {sup 99m}Tc occurs through shipping {sup 99}Mo samples to the medical facilities. The Neutron Source Facility of KIPT is the first ADS in the world operating at non-zero power and allows producing several isotopes for medical applications. The total power of the facility is about 300 kW, including ∼200 kW from fission reactions. The construction of the NSF has been already completed and operations will start during year 2016. After 20 irradiation days, {sup 82}Br, {sup 64}Cu, {sup 165}Dy, {sup 166}Ho, {sup 192}Ir, {sup 194}Ir, {sup 186}Re, {sup 188}Re and {sup 153}Sm have specific activities, larger than 100 MBq/mg without taking into account the self-shielding effect and therefore they are good candidates for production. The self-shielding effect plays an important role for all nuclide parents that have a large microscopic capture cross section. This effect reduces the reaction rate and therefore the specific activity. For instance, the analyses of molybdenum trioxide (MoO{sub 3}) irradiation samples have shown that the self-shielding effect may reduce the specific activity by a factor 1.4. The metallic form of natural molybdenum produces a higher specific activity, relative to the trioxide form. Analyses on the 6-day Ci {sup 99}Mo activity have shown that is possible to perform between 40 to 310 {sup 99m}Tc medical procedures per MoO{sub 3} irradiation week. The number of medical procedures depends on the diagnosed organ and the weight of the patient. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2024

Transfer functions for Q A /Q B international regulatory limits for the safe transport of radioactive materials

This paper presents a proposed revision of the International Atomic Energy Agency transport regulations, related to the A 1 and A 2 limit values used to determine the radioactive transport classification. Based on the 'Q system', a novel methodology was introduced to derive Q A and Q B values related to scenarios involving external exposure from a distant source. These values are key parameters that respectively represent the total effective dose and total equivalent dose to the skin, from all primary and secondary particles contributing to radiation exposure. The International Working Group (WG A 1 /A 2 ) is established and associated with the TRANSSC Technical Expert Group on Radiation Protection. A review of the A 1 and A 2 values is performed in response to identified limitations within the existing Q system. The followed approach is based on Monte Carlo simulations that enabled the development of transfer functions aimed at reducing computational time and increasing the flexibility of dose evaluations for any radionuclide with known particle emission spectra. This method allows updating the Q A and Q B values to account for future data evolutions (decay data, fluence-to-dose conversion coefficients) and standardizing the calculation of regulation limits across all referenced radionuclides and scenarios related to external exposure. The transfer functions are established using three Monte Carlo simulation codes—FLUKA, Geant4, and MCNP—and address the previous limitations of the 'Q system', reflecting the latest International Commission for Radiation Protection recommendations and improvements in calculation techniques. The results of the WG show consistent agreement across the codes, with minor discrepancies observed at low primary energies due to statistical uncertainties and different handling of stopping power for electrons/positrons in the codes. This revised approach aligns with current standards and recommendations, ensuring that the radiological consequences of transport accidents are acceptable for the new A 1 and A 2 limits from a radiological protection perspective.

61 RADIATION PROTECTION AND DOSIMETRY↗