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At least 181 records · Page 10

Two MCNP Models for Computational Performance Benchmarking

The purpose of this report is to describe two non-trivial MCNP models and execution configurations that are suitable to characterizing computational performance. To that end, this report uses a constructive solid geometry (CSG) representation of the Oak Ridge National Laboratory (ORNL) Pool Critical Assembly (PCA) [1–3] to perform a k-eigenvalue calculation and an unstructured mesh (UM) representation of the International Commission on Radiological Protection (ICRP) publication 145 (ICRP145) male human phantom [4, 5] to perform a fixed-source calculation assuming that a 1 MeV photon source is distributed throughout the phantom’s liver.

97 MATHEMATICS AND COMPUTING↗

National Criticality Experiments Research Center (NCERC) [Slides]

The National Criticality Experiments Research Center (NCERC) is a general-purpose critical assembly facility located within the Device Assembly Facility at the Nevada National Security Site. The NCERC is operated by Los Alamos National Laboratory.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Bench-Scale Electrolytic Dissolution of Quarter-Scale FCA Cans

In 2016, the Savannah River National Laboratory (SRNL) led, in support of and under sponsorship of the Department of Energy’s National Nuclear Security Administration (DOE/NNSA) Office of Material Management and Minimization (M3), the removal and transfer of the plutonium based Fast Critical Assembly (FCA) fuel from the Japan Atomic Energy Agency (JAEA) Tokai facility to the Savannah River Site (SRS). The team also included JAEA, multiple organizations in Savannah River Nuclear Solutions (SRNS), International Nuclear Services, and many other entities. The FCA fuel removal project completion was a key deliverable for M3 to the 2016 Nuclear Security Summit and constituted the largest inventory of weapons-usable plutonium removed under the nonproliferation program. The FCA materials consist of thousands of stainless steel (SS) clad plates and hundreds of SS clad rods. The FCA fuel elements were packaged in a carrier can and stored at SRS pending disposition of the fuel. Following an assessment of candidate disposition options, SRNS identified electrolytic dissolution (ED) as the most promising disposition option for the FCA plates and their preferred option was endorsed by DOE. This option entails electrochemically dissolving the entire FCA carrier can with fuel elements and was based on bench-scale laboratory testing and historical work on processing SS-clad and zirconium-clad uranium-based fuel in the H-Canyon electrolytic dissolver (last operated in 1980). The FCA plate consists of a plutonium-aluminum metal alloy core hermitically sealed in SS cladding.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Integral Experiment Request 305: CED-3b Summary Report

Under IER-305, critical experiments were done with and without molybdenum sleeves on 7uPCX fuel rods. The experiments were done in new critical assembly hardware designed to support the 7uPCX fuel in a 1.55 cm triangular-pitched array.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

(U) The Effect of ENDF/B-VIII.1 on Jezebel

The 239 Pu Jezebel critical assembly of 1954-55 was modeled in great detail and reevaluated using ENDF/B-VII.1 cross sections, and a new simplified (homogeneous one-dimensional spherical) benchmark was derived. The new simplified model was similar to that derived by Hansen and Paxton in 1969, but its radius was modified so that its k eff was the average of the C/E’s (where C is the calculated k eff and E is the inferred experimental k eff ) of the four detailed models. When ENDF/B-VIII.0 data were released, a mismatch between the detailed and simplified models was introduced due to changes in the nickel cross sections, because the detailed models contain nickel but the simplified model does not. A revision of the simplified model realigned the detailed and simplified models. ENDF/B-VIII.1 data have now been released. The simplified and detailed Jezebel models have been run using preliminary ENDF/B-VIII.1 ACE files. In Sec. II, it is shown that the detailed and simplified models are still aligned. No revision of the simplified model is called for. These results were computed using MCNP6.3.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

CSEWG Meeting Minutes

There were many improvements and additions to ENDF/B-VIII.1. We tested four key moderator materials: Be, HF, FLiBe, and graphite. Be-metal moderated critical assemblies of HEU only see 26 pcm difference. Sd library adds new physics to the TSL which improves the agreement with the benchmark. Other changes besides the TSL impact the benchmark: 235U and 9Be. Free gas change affects library heavily, hundreds of pcm in wrong way – need to look at FG libraries. F and Cl TSLs added to 8.1. A 3k pcm disagreement was seen without new TSL. Adding 8.1 + TSL reduces pcm disagreement A LOT.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Literature Review Investigating Historical Plutonium Solubility in SRS Tank Waste

The Savannah River Site (SRS) has designed the Accelerated Basin De-inventory (ABD) program to accelerate the de-inventory of L-Basin and accelerate the Spent Nuclear Fuel (SNF) disposition mission. Similarly, the H-Canyon facility at SRS is reestablishing the 6.3D electrolytic dissolver for dissolving unirradiated stainless-steel (SS) clad Fast Critical Assembly (FCA) fuel. In both discard types (ABD and FCA), plutonium is present and its complex solubility when composited to Concentration, Storage, and Transfer Facility (CSTF) sludge is being investigated as it may have downstream impacts to the liquid waste (LW) organization. This literature review aims to highlight and compile the existing literature on plutonium solubility in waste streams relevant to ABD and FCA discards, as well as discuss some considerations in analyzing solubility data of plutonium. This review serves to help define the analysis methods for future experiments involving plutonium (and other actinides) and in designing appropriate testing conditions surrounding these studies. This review is broken up into five parts and will discuss: (i) The possible effects of testing hold time and temperature on plutonium solubility, (ii) the influence of neutralization rate and particle size of freshly precipitated discards, (iii) the coprecipitation of plutonium with iron and uranium, (iv) predictive solubility modeling and the influence of supernate anions on solubility, and (v) the speciation of plutonium in solutionas a result of supernate anions.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Integral Experiments Request 441 CED-3b Summary Report

Under IER-441, critical experiments were done with and without tantalum test rods within a central test region surrounded by 7uPCX fuel rods. The experiments were done in new critical assembly hardware designed to support the 7uPCX fuel in a 1.02 cm triangular-pitched array. Appendix I is a draft of section 1 of the ICSBEP benchmark evaluation of the experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Integral Experiment Request 441 CED-4a Summary Report

Under IER-441, critical experiments were done with and without tantalum test rods within a central test region surrounded by 7uPCX fuel rods. The experiments were done in new critical assembly hardware designed to support the 7uPCX fuel in a 1.02 cm triangular-pitched array

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Evaluation of Pu Solubility in Glass for Sludge Batch 11

Stainless-steel clad Pu from Japan’s Fast Critical Assembly (FCA) reactor is currently being dispositioned at the Savannah River Site. The electrolytic dissolver, operated by Savannah River Nuclear Solutions in H-Canyon, is being utilized to dissolve the material. The resulting solutions are transferred to the Concentration, Storage, and Transfer Facilities, operated by Savannah River Mission Completion (SRMC), for subsequent vitrification at the Defense Waste Processing Facility (DWPF). In support of the FCA mission startup, a preliminary evaluation was conducted by the Savannah River National Laboratory to assess the impact of the FCA discards on the liquid waste system.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Gaseous fuel reactors for power systems

Gaseous-fuel nuclear reactors have significant advantages as energy sources for closed-cycle power systems. The advantages arise from the removal of temperature limits associated with conventional reactor fuel elements, the wide variety of methods of extracting energy from fissioning gases, and inherent low fissile and fission product in-core inventory due to continuous fuel reprocessing. Example power cycles and their general performance characteristics are discussed. Efficiencies of gaseous fuel reactor systems are shown to be high with resulting minimal environmental effects. A technical overview of the NASA-funded research program in gaseous fuel reactors is described and results of recent tests of uranium hexafluoride (UF6)-fueled critical assemblies are presented.

Kendall, J. S.↗

Automated inspection of solder joints for surface mount technology

Researchers at NASA/GSFC evaluated various automated inspection systems (AIS) technologies using test boards with known defects in surface mount solder joints. These boards were complex and included almost every type of surface mount device typical of critical assemblies used for space flight applications: X-ray radiography; X-ray laminography; Ultrasonic Imaging; Optical Imaging; Laser Imaging; and Infrared Inspection. Vendors, representative of the different technologies, inspected the test boards with their particular machine. The results of the evaluation showed limitations of AIS. Furthermore, none of the AIS technologies evaluated proved to meet all of the inspection criteria for use in high-reliability applications. It was found that certain inspection systems could supplement but not replace manual inspection for low-volume, high-reliability, surface mount solder joints.

Savage, Robert M.↗

Statistical Characterization of 18650-Format Lithium-Ion Cell Thermal Runaway Energy Distributions

Effective thermal management systems, designed to handle the impacts of thermal runaway (TR) and to prevent cell-to-cell propagation, are key to safe operation of lithium-ion (Li-ion) battery assemblies. Critical factors for optimizing these systems include the total energy released during a single cell TR event and the fraction of the total energy that is released through the cell casing vs. through the ejecta material. A unique calorimeter was utilized to examine the TR behavior of a statistically significant number of 18650-format Li-ion cells with varying manufacturers, chemistries, and capacities. The calorimeter was designed to contain the TR energy in a format conducive to discerning the fractions of energy released through the cell casing vs. through the ejecta material. Other benefits of this calorimeter included the ability to rapidly test of large quantities of cells and the intentional minimization of secondary combustion effects. High energy (270 Wh/kg) and moderate energy (200 Wh/kg) 18650 cells were tested. Some of the cells had an imbedded short circuit (ISC) device installed to aid in the examination of TR mechanisms under more realistic conditions. Other variations included cells with bottom vent (BV) features and cells with thin casings (0.22 1/4m). After combining the data gathered with the calorimeter, a statistical approach was used to examine the probability of certain TR behavior, and the associated energy distributions, as a function of capacity, venting features, cell casing thickness and temperature.

Thermal Runaway Energy Distribution↗

U-10Mo Monolithic Fuel Qualification Plan

The Material Management and Minimization (M3) Program’s primary objective within the U.S. Department of Energy/National Nuclear Security Administration is to achieve permanent threat reduction by minimizing and when possible, eliminating weapon-usable nuclear material around the world. Under M3’s Office of Conversion, the U.S. High Performance Research Reactor (USHPRR) Project, is pursuing a fuel qualification and licensing effort focused on converting high-performance research reactors in the United States from using highly enriched uranium (HEU) fuel to using low-enriched uranium (LEU) fuel. The body of this document is focused on defining the activities required for generic qualification of high density, LEU fuel that then allows reactor conversions to proceed for the four high-performance research reactors and the critical assembly that will operate using the U-10Mo monolithic fuel design (see Subsection 1.3): MITR, MURR, NBSR, ATR, and ATRC. Fuel qualification in this report means generic fuel acceptance of the U-10Mo monolithic fuel form by the NRC for use in reactor conversions in the United States, which allows the fuel to be used in subsequent reactor specific licensing requests. The U 10Mo Monolithic Fuel Research, Development, and Qualification Plan utilizes the functions and requirements of the USHPRR Project that was established in the Functions and Requirements Document (F&RD) [1] and expands on these requirements to ensure that planned tests have traceable results that will ensure the requirement has been met. The methods by which data will be collected to show that these requirements have been met are described in this document.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Preliminary MITR Design-Demonstration Element Thermal-Hydraulics and Structural Analyses

The United State High Performance Research Reactor (USHPRR) program aims to eliminate more than 200kg of High Enriched Uranium (HEU) from commerce annually by converting five U.S. high-performance research reactors and one associated critical assembly to Low Enriched Uranium (LEU) fuel using a high-density alloy of uranium-10 wt% molybdenum (U-10Mo). Massachusetts Institute of Technology Reactor (MITR) is one of five research reactors selected for this program. The objective of this report is to provide preliminary thermal-hydraulic and mechanical analyses of the hydrodynamic effects in the MITR Design Demonstration Element (DDE) under conservative approximations for the plate power distribution. This report provides details on the modeling approach and the simulation results obtained, including pressure, flow velocity, temperature, and oxide layer over the design demonstration experiment for the irradiation cycles in the Belgian Reactor (BR)-2.

42 ENGINEERING↗

Margin to Onset of Nucleate Boiling studies for MITR and NBSR Design Demonstration Experiments

The United State High Performance Research Reactor (USHPRR) program aims to eliminate more than 200kg of High Enriched Uranium (HEU) from commerce annually by converting five U.S. high-performance research reactors and one associated critical assembly to Low Enriched Uranium (LEU) fuel using a high-density alloy of uranium-10 wt% molybdenum (U-10Mo). The Massachusetts Institute of Technology Reactor (MITR) and the National Bureau of Standards Reactor (NBSR) are two of five research reactors selected for this program. Previous studies have addressed the thermal-hydraulics performance of the fuel Design Demonstration Elements (DDE) for these reactors under conservative assumptions in the Belgian Reactor (BR)-2. This report extends these previous studies by analyzing the margins to Onset of Nucleate Boiling (ONB) for these two DDEs under the conservative conditions used in the previous thermal-hydraulics analyses

42 ENGINEERING↗

Preliminary NBSR Design-Demonstration Element Thermal-Hydraulics and Structural Analyses

1.1 OBJECTIVE The United State High Performance Research Reactor (USHPRR) program aims to eliminate more than 200kg of High Enriched Uranium (HEU) from commerce annually by converting five U.S. high-performance research reactors and one associated critical assembly to Low Enriched Uranium (LEU) fuel using a high-density alloy of uranium-10 wt% molybdenum (U-10Mo). The National Bureau of Standards Reactor (NBSR) is one of five research reactors selected for this program. The objective of this report is to provide preliminary thermal-hydraulic and mechanical analyses of the hydrodynamic effects in the NBSR Design Demonstration Element (DDE) under conservative approximations for the plate power distribution. This report provides details on the modeling approach and the simulation results obtained, including pressure, flow velocity, temperature, and oxide layer over the design demonstration experiment for the irradiation cycles in the Belgian Reactor (BR)-2.

42 ENGINEERING↗