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At least 757 records · Page 42
All sourcesNASA NTRSDOE OSTI
DOE OSTI2013

APPLICATIONS OF CURRENT TECHNOLOGY FOR CONTINUOUS MONITORING OF SPENT FUEL

Advancements in technology have opened many opportunities to improve upon the current infrastructure surrounding the nuclear fuel cycle. Embedded devices, very small sensors, and wireless technology can be applied to Security, Safety, and Nonproliferation of Spent Nuclear Fuel. Security, separate of current video monitoring systems, can be improved by integrating current wireless technology with a variety of sensors including motion detection, altimeter, accelerometer, and a tagging system. By continually monitoring these sensors, thresholds can be set to sense deviations from nominal values. Then alarms or notifications can be activated as needed. Safety can be improved in several ways. First, human exposure to ionizing radiation can be reduced by using a wireless sensor package on each spent fuel cask to monitor radiation, temperature, humidity, etc. Since the sensor data is monitored remotely operator stay-time is decreased and distance from the spent fuel increased, so the overall radiation exposure is reduced as compared to visual inspections. The second improvement is the ability to monitor continuously rather than periodically. If changes occur to the material, alarm thresholds could be set and notifications made to provide advanced notice of negative data trends. These sensor packages could also record data to be used for scientific evaluation and studies to improve transportation and storage safety. Nonproliferation can be improved for spent fuel transportation and storage by designing an integrated tag that uses current infrastructure for reporting and in an event; tracking can be accomplished using the Iridium satellite system. This technology is similar to GPS but with higher signal strength and penetration power, but lower accuracy. A sensor package can integrate all or some of the above depending on the transportation and storage requirements and regulations. A sensor package can be developed using off the shelf technology and applying it to each specific need. There are products on the market for smart meters, industrial lighting control and home automation that can be applied to the Back End Fuel Cycle. With a little integration and innovation a cost effective solution is achievable.

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

Radioactive waste management approaches for developed countries

Nuclear power has demonstrated over the last 30 years its capacity to produce base-load electricity at a low, predictable and stable cost due to the very low economic dependence on the price of uranium. However the management of used nuclear fuel remains the “Achilles’ Heel” of this energy source since the storage of used nuclear fuel is increasing as evidenced by the following number with 2,000 tons of UNF produced each year by the 104 US nuclear reactor units which equates to a total of 62,000 spent fuel assemblies stored in dry cask and 88,000 stored in pools. Two options adopted by several countries will be presented. The first one adopted by Europe, Japan and Russia consists of recycling the used nuclear fuel after irradiation in a nuclear reactor. Ninety six percent of uranium and plutonium contained in the spent fuel could be reused to produce electricity and are worth recycling. The separation of uranium and plutonium from the wastes is realized through the industrial PUREX process so that they can be recycled for re-use in a nuclear reactor as a mixed oxide (MOX) fuel. The second option undertaken by Finland, Sweden and the United States implies the direct disposal of used nuclear fuel into a geologic formation. One has to remind that only 30% of the worldwide used nuclear fuel are currently recycled, the larger part being stored (70% in pool) waiting for scientific or political decisions. A third option is emerging with a closed fuel cycle which will improve the global sustainability of nuclear energy. This option will not only decrease the volume amount of nuclear waste but also the long-term radiotoxicity of the final waste, as well as improving the long-term safety and the heat-loading of the final repository. At the present time, numerous countries are focusing on the R&D recycling activities of the ultimate waste composed of fission products and minor actinides (americium and curium). Several new chemical extraction processes, such as TRUSPEAK, ALSEP, EXAM, or LUCA are pursued worldwide and their approaches will be highlighted.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Apparatus for safeguarding a radiological source

A tamper detector is provided for safeguarding a radiological source that is moved into and out of a storage location through an access porthole for storage and use. The radiological source is presumed to have an associated shipping container approved by the U.S. Nuclear Regulatory Commission for transporting the radiological source. The tamper detector typically includes a network of sealed tubing that spans at least a portion of the access porthole. There is an opening in the network of sealed tubing that is large enough for passage therethrough of the radiological source and small enough to prevent passage therethrough of the associated shipping cask. Generally a gas source connector is provided for establishing a gas pressure in the network of sealed tubing, and a pressure drop sensor is provided for detecting a drop in the gas pressure below a preset value.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2014

A Low-Tech, Low-Budget Storage Solution for High Level Radioactive Sources

The need for safe, secure, and economical storage of radioactive material becomes increasingly important as beneficial uses of radioactive material expand (increases inventory), as political instability rises (increases threat), and as final disposal and treatment facilities are delayed (increases inventory and storage duration). Several vendor-produced storage casks are available for this purpose but are often costly — due to the required design, analyses, and licensing costs. Thus the relatively high costs of currently accepted storage solutions may inhibit substantial improvements in safety and security that might otherwise be achieved. This is particularly true in areas of the world where the economic and/or the regulatory infrastructure may not provide the means and/or the justification for such an expense. This paper considers a relatively low-cost, low-technology radioactive material storage solution. The basic concept consists of a simple shielded storage container that can be fabricated locally using a steel pipe and a corrugated steel culvert as forms enclosing a concrete annulus. Benefits of such a system include 1) a low-tech solution that utilizes materials and skills available virtually anywhere in the world, 2) a readily scalable design that easily adapts to specific needs such as the geometry and radioactivity of the source term material), 3) flexible placement allows for free-standing above-ground or in-ground (i.e., below grade or bermed) installation, 4) the ability for future relocation without direct handling of sources, and 5) a long operational lifetime . ‘Le mieux est l’ennemi du bien’ (translated: The best is the enemy of good) applies to the management of radioactive materials – particularly where the economic and/or regulatory justification for additional investment is lacking. Development of a low-cost alternative that considerably enhances safety and security may lead to a greater overall risk reduction than insisting on solutions that remain economically and/or politically ‘out of reach’.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Systems for the Intermodal Routing of Spent Nuclear Fuel

The safe and secure movement of spent nuclear fuel from shutdown and active reactor facilities to intermediate or long term storage sites may, in some instances, require the use of several modes of transportation to accomplish the move. To that end, a fully operable multi-modal routing system is being developed within Oak Ridge National Laboratory s (ORNL) WebTRAGIS (Transportation Routing Analysis Geographic Information System). This study aims to provide an overview of multi-modal routing, the existing state of the TRAGIS networks, the source data needs, and the requirements for developing structural relationships between various modes to create a suitable system for modeling the transport of spent nuclear fuel via a multimodal network. Modern transportation systems are comprised of interconnected, yet separate, modal networks. Efficient transportation networks rely upon the smooth transfer of cargoes at junction points that serve as connectors between modes. A key logistical impediment to the shipment of spent nuclear fuel is the absence of identified or designated transfer locations between transport modes. Understanding the potential network impacts on intermodal transportation of spent nuclear fuel is vital for planning transportation routes from origin to destination. By identifying key locations where modes intersect, routing decisions can be made to prioritize cost savings, optimize transport times and minimize potential risks to the population and environment. In order to facilitate such a process, ORNL began the development of a base intermodal network and associated routing code. The network was developed using previous intermodal networks and information from publicly available data sources to construct a database of potential intermodal transfer locations with likely capability to handle spent nuclear fuel casks. The coding development focused on modifying the existing WebTRAGIS routing code to accommodate intermodal transfers and the selection of prioritization constraints and modifiers to determine route selection. The limitations of the current model and future directions for development are discussed, including the current state of information on possible intermodal transfer locations for spent fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Validation Study for Crediting Chlorine in Criticality Analyses for Spent Nuclear Fuel Disposition

Spent nuclear fuel (SNF) management practices in the United States rely on dry storage systems that include both canister- and cask-based systems. The United States Department of Energy Used Fuel Disposition Campaign is examining the feasibility of direct disposal of dual-purpose (storage and transportation) canisters (DPCs) in a geological repository. One of the major technical challenges for direct disposal is the ability to demonstrate the subcriticality of the DPCs loaded with SNF for the repository performance period (e.g., 10,000 years or more) as the DPCs may undergo degradation over time. Specifically, groundwater ingress into the DPC (i.e., flooding) could allow the system to achieve criticality in scenarios where the neutron absorber plates in the DPC basket have degraded. However, as was shown by Banerjee et al., some aqueous species in the groundwater provide noticeable reactivity reduction for these systems. For certain amounts of particular aqueous species (e.g., chlorine, lithium) in the groundwater, subcriticality can be demonstrated even for DPCs with complete degradation of the neutron absorber plates or a degraded fuel basket configuration. It has been demonstrated that chlorine is the leading impurity, as indicated by significant neutron absorption in the water that is available in reasonable quantities for the deep geological repository media under consideration. This paper presents the results of an investigation of the available integral experiments worldwide that could be used to validate DPC disposal criticality evaluations, including credit for chlorine. Due to the small number of applicable critical configurations, validation through traditional trending analysis was not possible. The bias in the eigenvalue of the application systems due only to the chlorine was calculated using TSURFER analysis and found to be on the order of 100 percent mille (1 pcm = 10 -5 k eff ). This study investigated the design of a series of critical configurations with varying amounts of chlorine to address validation gaps. Such integral experiments would support the crediting of the chlorine neutron-absorption properties in groundwater and the demonstration of subcriticality for DPCs in deep geologic repositories with sufficient chlorine availability.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Technical Basis for Peak Reactivity Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems

Oak Ridge National Laboratory and the United States Nuclear Regulatory Commission have initiated a multiyear project to investigate application of burnup credit for boiling-water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase (1) investigates applicability of peak reactivity methods currently used in spent fuel pools (SFPs) to storage and transportation systems and (2) evaluates validation of both reactivity (k eff ) calculations and burnup credit nuclide concentrations within these methods. The second phase will focus on extending burnup credit beyond peak reactivity. This paper documents the first phase, including an analysis of lattice design parameters and depletion effects, as well as both validation components. Initial efforts related to extended burnup credit are discussed in a companion paper. Peak reactivity analyses have been used in criticality analyses for licensing of BWR fuel in SFPs over the last 20 years. These analyses typically combine credit for the gadolinium burnable absorber present in the fuel with a modest amount of burnup credit. Gadolinium burnable absorbers are used in BWR assemblies to control core reactivity. The burnable absorber significantly reduces assembly reactivity at beginning of life, potentially leading to significant increases in assembly reactivity for burnups less than 15–20 GWd/MTU. The reactivity of each fuel lattice is dependent on gadolinium loading. The number of gadolinium-bearing fuel pins lowers initial lattice reactivity, but it has a small impact on the burnup and reactivity of the peak. The gadolinium concentration in each pin has a small impact on initial lattice reactivity but a significant effect on the reactivity of the peak and the burnup at which the peak occurs. The importance of the lattice parameters and depletion conditions are primarily determined by their impact on the gadolinium depletion. Criticality code validation for BWR burnup credit at peak reactivity requires a different set of experiments than for pressurized-water reactor burnup credit analysis because of differences in actinide compositions, presence of residual gadolinium absorber, and lower fission product concentrations. A survey of available critical experiments is presented along with a sample criticality code validation and determination of undercoverage penalties for some nuclides. The validation of depleted fuel compositions at peak reactivity presents many challenges which largely result from a lack of radiochemical assay data applicable to BWR fuel in this burnup range. In addition, none of the existing low burnup measurement data include residual gadolinium measurements. An example bias and uncertainty associated with validation of actinide-only fuel compositions is presented.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Effect of Control Blade History, and Axial Coolant Density and Burnup Profiles on BWR Burnup Credit

A technical basis for peak reactivity boiling water reactor (BWR) burnup credit (BUC) methods was recently generated, and the technical basis for extended BWR BUC is now being developed. In this paper, a number of effects related to extended BWR BUC are analyzed, including three major operational effects in BWRs: the coolant density axial distribution, the use of control blades during operation, and the axial burnup profile. Specifically, uniform axial moderator density profiles are analyzed and compared to previous results and an additional temporal fidelity study combing moderator density profiles for three different fuel assemblies is presented. Realistic control blade histories and cask criticality results are compared to previously generated constructed control blade histories. Finally, a preliminary study of the axial burnup profile is provided.

Marshall, William BJ J↗
DOE OSTI2017

Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS)

This will be licensed via RSICC. A new, integrated data and analysis system has been designed to simplify and automate the performance of accurate and efficient evaluations for characterizing the input to the overall nuclear waste management system -UNF-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS). A relational database within UNF-ST&DARDS provides a standard means by which UNF-ST&DARDS can succinctly store and retrieve modeling and simulation (M&S) parameters for specific spent nuclear fuel analysis. A library of various analysis model templates provides the ability to communicate the various set of M&S parameters to the most appropriate M&S application. Interactive visualization capabilities facilitate data analysis and results interpretation. UNF-ST&DARDS current analysis capabilities include (1) assembly-specific depletion and decay, (2) and spent nuclear fuel cask-specific criticality and shielding. Currently, UNF-ST&DARDS uses SCALE nuclear analysis code system for performing nuclear analysis.

Banerjee, Kaushik↗
DOE OSTI2015

Remanent Activation in the Mini-SHINE Experiments

Argonne National Laboratory is assisting SHINE Medical Technologies in developing a domestic source of the medical isotope 99Mo through the fission of low-enrichment uranium in a uranyl sulfate solution. In Phase 2 of these experiments, electrons from a linear accelerator create neutrons by interacting in a depleted uranium target, and these neutrons are used to irradiate the solution. The resulting neutron and photon radiation activates the target, the solution vessels, and a shielded cell that surrounds the experimental apparatus. When the experimental campaign is complete, the target must be removed into a shielding cask, and the experimental components must be disassembled. The radiation transport code MCNPX and the transmutation code CINDER were used to calculate the radionuclide inventories of the solution, the target assembly, and the shielded cell, and to determine the dose rates and shielding requirements for selected removal scenarios for the target assembly and the solution vessels.

Micklich, Bradley J.↗
DOE OSTI2018

Detailed SCALE Dose Rate Evaluations for a Consolidated Interim Spent Nuclear Fuel Storage Facility

This paper presents a confirmatory dose rate calculation for a proposed consolidated interim spent nuclear fuel (SNF) storage facility (CISF). The proposed CISF was modeled in detail using SCALE 6.2.1 to determine the radiation dose map of the surrounding areas of the facility and determine the location of the controlled area boundary based on 10 CFR 72.104 requirements. The spatial dose rate distribution was determined as the total dose rate contributions by each of the 467 storage casks comprising the CISF. A rectangular boundary outside of which no dose rate exceeds 25 mrem/yr has the dimensions of 1.25 km x 1.25 km. It denotes the smallest rectangle that could serve as the controlled area boundary. The results (dose map) show the unique capability of the SCALE code system for a high-resolution radiation dose simulation of a large SNF storage facility.

Radulescu, Georgeta↗
DOE OSTI2018

Using Deep Machine Learning to Conduct Object-Based Identification and Motion Detection on Safeguards Video Surveillance

Video surveillance is one of the core monitoring technologies used by the International Atomic Energy Agency (IAEA) Department of Safeguards at safeguarded nuclear facilities worldwide. Current IAEA image-review software has functions for scene-change detection, black image detection and missing scene analysis, but their capabilities are not optimum. The current workflow for the detection of safeguards relevant events heavily depends on inspectors’ laborious visual examination of surveillance videos, which is a time-consuming process and prone to errors. To improve the accuracy of the process and reduce inspectors’ burden, the paper proposes using deep machine learning to detect objects of interest in video streams and to conduct object-based motion detection. The hypothesis of this work is that deep machine learning will reduce the burden on inspectors and reduce errors by automatically locating and identifying objects of interest in video streams. Objects of interest include casks and fuel assemblies that are typically monitored by inspectors. The algorithm being developed in this work is based on a computationally efficient deep machine learning algorithm – You Only Look Once (YOLO) – but is further devised to address specific challenges related to the operation of nuclear facilities. The developed model (which is called YOLO-NSG – YOLO for Nuclear Safeguards) is evaluated with data sets collected at the test facilities at Brookhaven National Laboratory (BNL) and Sandia National Laboratories (SNL). The initial focus of the research is for application at safeguarded nuclear reactors, such as pressurized heavy-water reactors, where video surveillance is broadly deployed, but can be extended to other use cases of nuclear safeguards. The detailed structure of YOLO-NSG model is introduced and the test results are reported in the paper.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Data Analysis and Detector Troubleshooting for the Silicon Muon Scanner

Muon tomography is the calculation of the density distribution of a material from the probable angles of muon Coulomb scattering in a sampled object. This allows the probing of internal object structures that cannot otherwise be observed. It has many important applications for both national security and other fields, such as archaeology and geology. Its uses range from detecting radioactive material intended for nuclear weaponry to searching for hidden chambers in the Egyptian pyramids. In this project we intend to demonstrate the capability of a tomographic system which uses cosmic rays as the muon source and tracks each muon's initial and final trajectories to ascertain information about their scattering pattern. This method can be used to determine the structure or composition of an object, such as testing to see if a nuclear fuel cask is empty or full. We intend this as an early R&D exploration into the practicality of implementing silicon microstrip sensors in border scans for radioactive contraband, as an alternative to existing drift tube technology.

43 PARTICLE ACCELERATORS↗
DOE OSTI2014

Sensitivity Studies on UNF Cladding Structural Integrity under NCT Loading

This work describes the sensitivity analyses performed on the transportation package used in the NCT initial modeling and simulation demonstration [1]. The initial demonstration predicted the structural integrity of the fuel cladding by integrating models from the fuel rod/grid slot level, the single assembly-level, and the full cask-level. A number of sensitivity analyses were performed on the models at the different levels to determine how sensitive the results were to the parameters that were varied. The results are used to coarsely group the parameters investigated into four categories of sensitivity: No Sensitivity, Low Sensitivity; Moderate Sensitivity, and High Sensitivity. The High Sensitivity analyses are described in detail below. The Moderate, Low and No Sensitivity results are briefly described.

used nuclear fuel, UNF, normal conditions of trans↗
DOE OSTI2019

History and Status of DOE’s Standardized Canister

[Poster] The U.S. Department of Energy (DOE) is responsible for managing spent nuclear fuel (SNF) currently in its possession, as well as SNF that will come into its possession in the future. DOE will rely upon the standardized canister to confine radionuclides and precluding the moderator in order to ensure safe, long-term management and disposition of SNF in its possession. This strategy helps minimize characterization costs (i.e., funds and exposure) and other technical challenges associated with the traditional approach of relying on fuel-specific properties to demonstrate compliance with the regulator. The use of the standardized canister offers the following benefits: • The standardized canister design is complete. It has been tested under a variety of drop and accident scenarios and was subsequently licensed under 10 CFR 72 as a storage system as part of the Foster Wheeler Environmental Corporation (FWENC) Facility. It is also part of the license submittal for Yucca Mountain. • There is high confidence that the standardized canister design can be stored within commercially-available storage casks and licensed as part of a transportation-system package. • A robust standardized canister simplifies accident analyses and does not rely on characterization of the contained waste. This simplifies safety analysis, provides a greater level of confidence with stakeholders and reduces cost and risk. • A common design for the associated packaging, closure, and loadout facilities results in significant savings. • The DOE standardized canister would simplify design efforts, procurement, and fabrication. • Standardizing handling activities and associated training, procedures, equipment, and tooling leads to enhanced safety and surety of operations. The standardized canister is the preferred package for these materials. It withstood scrutiny by the Yucca Mountain Repository License Application, the Nuclear Regulatory Commission’s (NRC) Advisory Committee on Nuclear Waste, the Nuclear Waste Technical Review Board, an American Society of Mechanical Engineers peer review, and the FWENC storage license application. This evaluation summarizes and documents the status of the design and rationale for the DOE standardized canister, and outlines remaining needs before deployment. In addition, it discusses the factors that have the potential to influence storage, transportation, and disposal of DOE SNF.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

History and Status of DOE’s Standardized Canister – 19657

The U.S. Department of Energy (DOE) is responsible for managing spent nuclear fuel (SNF) that is currently in, or will in the future come into, its possession. The DOE standardized canister will be relied upon to confine radionuclides and preclude moderator in order to ensure safe, long-term management and disposition of DOE SNF. This strategy offers an avenue for avoiding the characterization costs (i.e. funds and exposure) and other technical challenges associated with the traditional approach of relying on fuel-specific properties as the basis for demonstrating compliance with the regulator. The use of the standardized canister offers the following benefits: • The standardized canister design is complete. It has been tested under a variety of drop and accident scenarios and was subsequently licensed under 10 CFR 72 as a storage system as part of the FWENC Facility. It is also part of the license submittal for Yucca Mountain. • There is high confidence that the standardized canister design can be stored within commercially available storage casks and licensed as part of a transportation-system package. • The robustness of the standardized canister simplifies accident analyses and does not rely on characterization of the contained waste. This simplifies safety analysis, provides a greater level of confidence with stake holders and reduces cost and risk. • Significant savings result from a common design for the associated packaging, closure, and loadout facilities. • Design efforts, procurement, and fabrication are simplified • Safety and surety of operations are enhanced by standardizing handling activities and associated training, procedures, equipment, and tooling. The standardized canister is the preferred package for these materials and has withstood scrutiny by the Yucca Mountain Repository License Application, the NRC’s Advisory Committee on Nuclear Waste, the Nuclear Waste Technical Review Board, an American Society of Mechanical Engineers peer review, and the Foster Wheeler Environmental Corporation storage license application. This evaluation summarizes and documents the status of the design and rationale for the DOE standardized canister and outlines the remaining needs to allow deployment. In addition, it discusses the factors that have the potential to influence storage, transportation, and disposal of DOE SNF.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

A Misload Analysis Methodology Supporting Criticality Analysis of Spent Nuclear Fuel Canisters Using As-Loaded Configurations

This paper presents an assembly misload analysis methodology developed to support criticality safety analysis of spent nuclear fuel (SNF) dual-purpose canisters (DPCs) using as-loaded configurations. The misload analysis approach is based on Interim Staff Guidance 8 “Burnup Credit in the Criticality Safety Analyses of PWR Spent Fuel in Transportation and Storage Casks” (ISG 8 rev. 3), but extended to support as-loaded configurations. The two misload scenarios analyzed included; (1) the correct assemblies are selected from the pool but placed incorrectly into the most reactive configuration inside the canister, and (2) the incorrect, most reactive assembly/assemblies in the pool is/are placed into the most reactive position(s) in the canister. Results from this misload analysis approach can be combined with misload probability to support criticality safety assessment of SNF during storage, transportation, and disposal. The misload analysis has been applied to 67 loaded DPCs at two sites using a disposal scenario. The analyzed sites include both PWR and BWR reactors and two canister variants—MPC-32 and MPC-68.

Liljenfeldt, Henrik↗
DOE OSTI2020

As-run physics analysis for the EPRI-3 experiment for cycles 155B and 158B

The purpose of this Engineering Calculations Analysis Report (ECAR) is to document the as-run heat generation rates (W/g), flux/fluence, radio-isotopic source term in Curies and grams, decay heat rate in Watts, and specimen DPA for the post-irradiation shipment and examination of the Electric Power Research Institute (EPRI)-3-1 and EPRI-3-2 experiments. EPRI-3-1 was irradiated during cycle 155B ending April 12, 2014. EPRI-3-2, a modification of EPRI-3-1, was irradiated during cycle 158B ending April 1, 2016. Source term, decay heat rate, and DPA calculations were based on scaled Monte Carlo N-Particle (MCNP) calculated fluxes for the EPRI-3-2 experiment. Although EPRI-3-2 is a modification of EPRI-3-1, no significant changes were made in the placement of experiment specimens. Calculated fluxes were scaled using as-run center lobe source powers of the corresponding irradiation cycles for EPRI-3-1 and EPRI-3-2. ORIGEN2 Version 2.2 was used to determine the decay heat rate and source term for the EPRI-3 experiment. The source term analysis was performed as requested by the project staff. The results of this analysis may be used to demonstrate compliance with shipping requirements in the BEA Research Reactor (BRR) cask following irradiation as well as to provide as-run source term to support Post-Irradiation Examination (PIE).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗