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At least 1,801 records · Page 100

Mesos-scale modeling of irradiation in pressurized water reactor concrete biological shields

Neutron irradiation exposure causes aggregate expansion, namely radiation-induced volumetric expansion (RIVE). The structural significance of RIVE on a portion of a prototypical pressurized water reactor (PWR) concrete biological shield (CBS) is investigated by using a meso- scale nonlinear concrete model with inputs from an irradiation transport code and a coupled moisture transport-heat transfer code. RIVE-induced severe cracking onset appears to be triggered by the ini- tial shrinkage-induced cracking and propagates to a depth of > 10 cm at extended operation of 80 years. Relaxation of the cement paste stresses results in delaying the crack propagation by about 10 years.

concrete↗

Characterization of Radiation Fields in Biological Shields of Nuclear Power Plants for Assessing Concrete Degradation

Life extensions of nuclear power plants to 60 and potentially 80 years of operation have renewed interest in long-term material degradation. One material being considered is concrete with a particular focus on radiation-induced effects. Based on the projected neutron fluence (E > 0.1 MeV) values in the concrete biological shields of the US PWR fleet and the available data on radiation effects on concrete, some decrease in mechanical properties of concrete cannot be ruled out during extended operation beyond 60 years. An expansion of the irradiated concrete database and a reliable determination of relevant neutron fluence energy cutoff value are necessary to assure reliable risk assessment for NPPs extended operation.

Remec, Igor↗

Lightweight concrete with enhanced neutron shielding

A lightweight concrete containing polyethylene terephthalate in an amount of 20% by total volume. The concrete is enriched with hydrogen and is therefore highly effective at thermalizing neutrons. The concrete can be used independently or as a component of an advanced neutron radiation shielding system.

36 MATERIALS SCIENCE↗

Energy Deposition and Shielding Study of the Front End for the Neutrino Factory

In the Neutrino Factory and Muon Collider muons are produced by firing high energy protons onto a target to produce pions. The pions decay to muons which are then accelerated. This method of pion production results in significant background from protons and electrons, which may result in heat deposition on superconducting materials and activation of the machine preventing manual handling. In this paper we discuss the design of a secondary particle handling system. The system comprises a solenoidal chicane that filters high momentum particles, followed by a proton absorber that reduces the energy of all particles, resulting in the rejection of low energy protons that pass through the solenoid chicane. We detail the design and optimization of the system and energy deposition and shielding analysis in MARS15.

43 PARTICLE ACCELERATORS↗

Characterization of Radiation Fields in Biological Shields of Nuclear Power Plants for Assessing Concrete Degradationa

Life extensions of nuclear power plants to 60 and potentially 80 years of operation have renewed interest in long-term material degradation. One material being considered is concrete, with a particular focus on radiation-induced effects. Based on the projected neutron fluence values (E > 0.1 MeV) in the concrete biological shields of the US pressurized water reactor fleet and the available data on radiation effects on concrete, some decrease in mechanical properties of concrete cannot be ruled out during extended operation beyond 60 years. An expansion of the irradiated concrete database and a reliable determination of relevant neutron fluence energy cutoff value are necessary to ensure reliable risk assessment for extended operation of nuclear power plants.

concrete↗

Electrostatic shielding of transformers

Toroidal transformers are currently used only in low-voltage applications. There is no published experience for toroidal transformer design at distribution-level voltages. Toroidal transformers are provided with electrostatic shielding to make possible high voltage applications and withstand the impulse test.

24 POWER TRANSMISSION AND DISTRIBUTION↗

Shield devices, systems, and methods for improved measurements and detection

The present disclosure relates to a device having a first wall that includes a first opening having a first edge, and a first aperture through the first wall. The first edge may be positioned at least partially within a first plane, the first wall and the first plane may define a first space positioned within the first wall, and the first aperture may have a second edge with at least a portion of the second edge positioned outside of the first plane. The device may be utilized to shield one or more sensors from radiation, to insure measurements made by the one or more sensors more accurately represent the actual environmental conditions.

Lustbader, Jason A.↗

Heat shield for pressure casing

A compressor may include a casing defining a discharge cavity and a seal cavity. A rotary shaft may be disposed in the casing, and a shaft seal assembly may be disposed in the seal cavity and about the rotary shaft. An impeller may be coupled with and configured to be driven by the rotary shaft. A balance piston may be integral with the impeller and may define the discharge cavity and the seal cavity. A balance piston seal may be disposed about the balance piston such that the balance piston seal and the balance piston define a radial clearance therebetween. The radial clearance may be configured to provide fluid communication from the impeller to the discharge cavity. A heat shield may be disposed in the discharge cavity, and may be configured to prevent the conduction of heat from the discharge cavity to the seal cavity via the casing.

42 ENGINEERING↗

Lid actuation system for shielded cask

A lid system for a shielding cask is provided, the system comprising a plurality of contiguous panels, each panel is axially actuated by a first lift piston located laterally from the casks' longitudinal principal axis, and rotated about the cask by a ‘hinge’ piston at approximately 90 degrees along an arc from the first piston, the arc formed by a portion of the lip of the cask.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Modified eddy current probe having a faraday shield

A modified eddy current (MEC) probe includes a probe body having a bore formed therein. A printed circuit board (PCB) assembly includes a circuit board defining a plane, has a plurality of electronic components mounted thereon, and is configured for mounting within the bore of the probe body. A coil board assembly is electrically connected to the PCB assembly, defines a plane, and includes a first coil board layer, a second coil board layer bonded to the first coil board layer, a transmitter coil formed on an outside surface of the first coil board layer, a Faraday shield formed on an outside surface of the second coil board layer, and a sensor coil formed on one of an inside surface of the first coil board layer and the second coil board layer, such that the sensor coil is positioned between the first coil board layer and the second coil board layer. The plane of the coil board assembly is arranged orthogonally to the plane of the circuit board.

47 OTHER INSTRUMENTATION↗

Comparison of Attila Transport Code with MCNP in the Analysis of Small Modular Reactor Shielding

Of great importance in the design of a nuclear power plant is the evaluation of shielding from the ionizing radiation produced in the core of the nuclear reactor. This requires simulation software to predict the dose rates in the areas of interest. These problems are referred to as deep penetration problems, as they involve the reduction of a radiation field by two or three orders of magnitude or more. This makes accurate analyses difficult for Monte Carlo based solutions that rely on enough particles being sampled in an area of interest. Comparing the simulation of the same modeled system with more than one computer application, especially one that uses a different algorithmic approach, is a way to gain confidence in the results. In this research, two models were created, one in Monte Carlo N Particle (MCNP) which uses a stochastic algorithmic approach, and the other in Attila, which uses a deterministic algorithmic approach to test the validity of the model and to estimate the dose rates in the area of interest. Successful comparison of the two models is a way we can validate the model used for that estimation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Radiation Shielding Analysis for the PIP-II Linac at Fermilab

The Proton Improvement Plan-II (PIP-II) [1] has been developed at Fermilab to provide powerful proton beams to the laboratory’s experiments. An 800-MeV superconducting linear accelerator—a centerpiece of the project—is currently under construction in Batavia, Illinois (USA). After completion, the superconducting linac will be the starting point for the 1.2 MW (Phase 1) and 2.4-MW (Phase 2) proton beam that is needed for the Long-Baseline Neutrino Facility (LBNF) at Fermilab [2]. Due to unavoidable loss of a fraction of the beam in the accelerator components, a certain level of radiation will be generated in the accelerator tunnel both during normal operation and at accidents. This work deals with radiation shielding design for the accelerator facility.

Rakhno, Igor↗

Systems and methods for shielding falling particles within a solar thermal falling particle receiver

Systems and methods for falling particle receivers are disclosed that include shield or deflector structures around the receiver aperture to reduce wind effects and/or heat losses from the falling particles. External and internal structures are disclosed that can be tailored to reduce particle, thermal, and radiative losses from within the cavity receiver due to external wind and the falling particles that are irradiated within the receiver. Structures of varying shapes, sizes, and composition (transparent, reflective) are described.

14 SOLAR ENERGY↗

Use of a Shielded High Resolution Gamma Spectrometry System to Segregate LLW from Contact Handleable ILW Containing Plutonium - 13046

Dounreay Site Restoration Limited (DSRL) have a number of drums of solid waste that may contain Plutonium Contaminated Material. These are currently categorised as Contact Handleable Intermediate Level Waste (CHILW). A significant fraction of these drums potentially contain waste that is in the Low Level Waste (LLW) category. A Canberra Q2 shielded high resolution gamma spectrometry system is being used to quantify the total activity of drums that are potentially in the LLW category in order to segregate those that do contain LLW from CHILW drums and thus to minimise the total volume of waste in the higher category. Am-241 is being used as an indicator of the presence of plutonium in the waste from its strong 59.54 keV gamma-ray; a knowledge of the different waste streams from which the material originates allows a pessimistic waste 'fingerprint' to be used in order to determine an upper limit to the activities of the weak and non-gamma-emitting plutonium and associated radionuclides. This paper describes the main features of the high resolution gamma spectrometry system being used by DSRL to perform the segregation of CHILW and LLW and how it was configured and calibrated using the Canberra In-Situ Object Counting System (ISOCS). It also describes how potential LLW drums are selected for assay and how the system uses the existing waste stream fingerprint information to determine a reliable upper limit for the total activity present in each measured drum. Results from the initial on-site commissioning trials and the first measurements of waste drums using the new monitor are presented. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Use of Ductile Iron for Radioactive Waste Packaging and Shielding in the United States - 17278

Ductile Cast Iron (DCI) has been used for packaging radioactive waste in Europe and the United Kingdom for many years. The ductile iron provides many favorable properties when compared with steel and offers many advantages including: - Monolithic structure increasing overall strength; - No welding required; no tempering or maintenance/inspection of the weld area is required; - Fast, serial production of similar containers reducing manufacturing costs; - Lower cost compared to steel or steel/lead composites; - Good corrosion resistance even when uncoated; - Can be coated or plated to provide suitable surface for decontamination; - Flexibility in design due to the manufacturing process permits a variety of shapes and very large size container possibilities; - Hazardous materials not needed or used for shielding. The flexibility in design will help generators find viable and more cost effective packaging options for radioactive waste. The package is conformed to the waste instead of conforming waste to the package. Commercial utilities often replace radioactive components which are too large for conventional packaging and have dose rates too high for flexible wraps while plant schedule and physical configuration won't allow for component segmentation. This is one situation where the use of ductile iron can be used to provide a packaging solution. By assessing the radiological characterization and component size and weight, ductile iron can be cast into certified containers to easily package, ship and dispose of the waste in a cost effective manner. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Development of a Type A Shielded Liquid Package - 18215

In the U.S., very few cost-effective certified packages exist for shipping radioactive Type A quantities of liquids. Two activities where the lack of such a package was identified as creating a serious need were: 1) shipping Mo-99, both across the United States for medical purposes; as well as, 2) shipping high Cs-137 concentration High Level Waste (HLW) supernatant from the Savannah River Site and/or the Hanford Site potentially on public roads and highways for detailed laboratory characterization. For shipping Mo-99, the need for a package was amplified by its ubiquitous use in the radio-medical industry. Although Tc-99m is used in approximately 80% of the medical radioisotope procedures, it is almost universally shipped as Mo-99, with the Mo-99 then converted to Tc-99m near the point of use (e.g. at the doctor's office or hospital). This transport of Mo-66 with the subsequent conversion to Tc- 99m is because Mo-99 has a 66-hour half-life, while Tc-99m has only a 6-hour half-life. With the noted need for packaging to support Mo-99 shipments, a new cost-effective drum based packaging which could transport Type A quantities of liquids (including those with a high specific activity/gamma energy) for use on public highways and commercial aircraft was developed. Starting with the well analyzed design and analysis of a physically similar 'inner 30 gallons, outer 50 gallons' packaging (i.e. DOE 9979 package) certified for Type AF quantities of solids, this paper discusses the additional design, regulatory requirements, and tests associated in certifying the Heavily-shielded Type A Liquid-package (HVTAL). Based on analyses, the authorized content of the package was then expanded for additional Type A non-fissile liquids, including HLW supernatant. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

SCALE 6.3.1 Radiation Source Terms and Shielding Analysis for a Postulated Sodium-Cooled Fast Reactor Accident Scenario

In support of the US Nuclear Regulatory Commission non–light-water reactor fuel cycle demonstration project, SCALE 6.3.1 capabilities for radiation source term and shielding calculations are demonstrated for scenarios in the sodium-cooled fast reactor (SFR) fuel cycle. A postulated accident scenario, which consists of a seismic event causing the refueling machine to fall and release a spent fuel assembly inside the containment building (CB), is analyzed in this paper. Radiation source terms were generated for a U/TRU-10Zr metal fuel assembly with a 16.5% initial transuranic waste content and a discharge burnup of approximately 95 GWd/tHM; source terms were also generated for a high-assay low-enriched uranium metal fuel assembly (U-10Zr) with a 16.5% initial enrichment and a discharge burnup of 149.74 GWd/tHM. These radiation source terms were then used to determine the dose rate inside the CB and near the outer surface of the CB for a range of spent fuel assembly cooling times. The dose rate produced by the analyzed SFR assemblies is similar to that produced by a typical pressurized water reactor assembly with a discharge burnup of 50 GWd/MTU. Ultimately, the validation of the source terms predicted for SFRs with SCALE will need to be demonstrated via the use of assay measurements.

Radulescu, Georgeta↗

Overview of Shielding Analyses at Oak Ridge National Laboratory’s Spallation Neutron Source Second Target Station

The Neutronics Group of the Second Target Station project at Oak Ridge National Laboratory is responsible for all neutronics analyses related to design and construction. This paper provides an overview of four neutronics analyses performed for the Second Target Station project, which is representative of all the ongoing work but especially tasks related to shielding. These analyses extend from the proton accelerator through the target monolith and bunker to the end of a neutron beamline. Each example highlights the tools and methods the Neutronics Group uses for analysis. The primary tool is the Monte Carlo radiation transport code MCNP 6.2, but this is augmented by other codes that supplement the input of MCNP. The other codes specifically highlighted in this paper include Attila4MC, ADVANTG, and AARE.

Miller, Thomas↗

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