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All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

The pursuit of net-positive sustainability for industrial decarbonization with hybrid energy systems

We report signatories of the Paris Agreement are set to miss their climate targets. The net-zero pledges announced to date across many countries and private industries are insufficient to achieve carbon neutrality, which requires implementation of far-reaching and significantly scaled-up climate-positive actions. All technology options that pertain to deep decarbonization and carbon removal must be part of the mitigation portfolio. Transformative action plans must be established in which every individual/organization around the globe is an actor of changes in relation to net-zero goals. Such plans must involve governmental policy support but also encourage voluntary efforts to boost innovations, investments, initiatives, and behavioral changes. Recognition of these efforts is made quantifiable with the concepts of carbon handprint and net positivity. This paper presents a carbon handprint perspective on characterizing the environmental benefits of hybrid energy systems (HESs)—a widely applicable solution to cleaner production leveraging the capabilities of a portfolio of low-carbon energy sources—that provide heat and electricity to industrial processes. First, the carbon handprint and net positivity concepts and their calculation approach are introduced. The state of the art of HES-enabled industrial cogeneration is then surveyed, and the greenhouse gas emission intensities of different energy sources that power HESs are compared. Next, drawing on a case study about a US chemical facility's voluntary initiative to explore replacing its fossil fuel–based cogeneration infrastructure with a clean energy–generated HES, several technically viable scenarios are evaluated—especially those with small modular nuclear reactors—to illustrate how the positive-thinking handprint approach helps encourage and inform the search for widespread influence pathways in pursuit of net-positive sustainability. Finally, current knowledge gaps in the case study are identified, and opportunities to scale up the proposed handprint-based analysis are outlined with consideration of an expanded role of HESs in fulfilling climate objectives.It is envisioned that like-minded decision makers in the industry sector and beyond will adopt this perspective and act synergistically to enhance their environmental stewardship through voluntary actions and make greater contributions to the planet's climate future.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
NASA NTRS2017

Kilopower: Small and Affordable Fission Power Systems for Space

The Nuclear Systems Kilopower Project was initiated by NASA's Space Technology Mission Directorate Game Changing Development Program in fiscal year 2015 to demonstrate subsystem-level technology readiness of small space fission power in a relevant environment (Technology Readiness Level 5) for space science and human exploration power needs. The Nuclear Systems Kilopower Project centerpiece is the Kilopower Reactor Using Stirling Technology (KRUSTY) test, which consists of the development and testing of a fission ground technology demonstrator of a 1 kWe-class fission power system. The technologies to be developed and validated by KRUSTY are extensible to space fission power systems from 1 to 10 kWe, which can enable higher power future potential deep space science missions, as well as modular surface fission power systems for exploration. The Kilopower Project is cofounded by NASA and the Department of Energy National Nuclear Security Administration (NNSA).KRUSTY include the reactor core, heat pipes to transfer the heat from the core to the power conversion system, and the power conversion system. Los Alamos National Laboratory leads the design of the reactor, and the Y-12 National Security Complex is fabricating it. NASA Glenn Research Center (GRC) has designed, built, and demonstrated the balance of plant heat transfer and power conversion portions of the KRUSTY experiment. NASA MSFC developed an electrical reactor simulator for non-nuclear testing, and the design of the reflector and shielding for nuclear testing. In 2016, an electrically heated non-fissionable Depleted Uranium (DU) core was tested at GRC in a configuration identical to the planned nuclear test. Once the reactor core has been fabricated and shipped to the Device Assembly Facility at the NNSAs Nevada National Security Site, the KRUSTY nuclear experiment will be assembled and tested. Completion of the KRUSTY experiment will validate the readiness of 1 to 10 kWe space fission technology for NASAs future requirements for sunlight-independent space power. An early opportunity for demonstration of In-Situ Resource Utilization (ISRU) capability on the surface of Mars is currently being considered for 2026 launch. Since a space fission system is the leading option for power generation for the first Mars human outpost, a smaller version of a planetary surface fission power system could be built to power the ISRU demonstration and ensure its end-to-end validity. Planning is underway to start the hardware development of this subscale flight demonstrator in 2018.

Space nuclear power↗
DOE OSTI2021

MOOSE Framework Meshing Enhancements to Support Reactor Analysis

MOOSE-based physics codes require an input finite element mesh on which the physics solution is calculated, reported, and transferred to other physics codes. The use of difficult-touse, external licensed software is often required to generate high quality meshes for reactor geometries. High-fidelity geometry modeling also requires elaborate tracking of groups of elements for material property assignment and output reporting which can be considerably complex for the user to identify and maintain. Under the U.S. Department of Energy Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, several meshingrelated enhancements have been developed for the MOOSE framework to address user challenges in creating finite element meshes for advanced reactor geometries. MOOSE mesh generators have been developed to mesh hexagonal geometries (pins, ducted assemblies, and cores) commonly found in liquid-metal cooled fast reactor concepts. The mesh generator used for hexagonal pin cells is generic for regular polygons and therefore may also be used for Cartesian pin cells. Hexagonal pin cells can be stitched into ducted assemblies, and assemblies can be stitched together into a core. The user may specify region ids, region names, and other preferences on the mesh. This control is useful for later material mapping in the MOOSE-based physics codes input. A capability was also developed for meshing rotating control drums including determination of material volume fractions in each mesh element as a function of time. Control drum meshes may be stitched to other hexagonal assemblies to create a core configuration. Additional mesh generators were developed that wrap around the hexagonal meshing capabilities and utilize “extra element integer” ID values on each element. In regular Cartesian or hexagonal assemblies or cores, the bookkeeping of element groups for both material assignment and output reporting can now be automated through assignment of pin, assembly, core, axial and depletion id values stored as extra element integers. The extra element tags on the mesh greatly speed the reactor analyst’s efforts to map materials to meshes, track depletion zones, and parse output such as axial pin power distributions. At the highest level, pin, assembly, and core mesh generators (with this reactor terminology) have also been developed to easily generate regular Cartesian and hexagonal cores, including axial extrusion. These reactor geometry builders call upon the previously mentioned capabilities to produce analysis-ready 3D meshes including material assignments. Open source mesh triangulation capabilities were also investigated for integration into the MOOSE framework to address the need for meshing the core periphery region which extends from the irregular outer assembly border to a cylindrical boundary. Options are limited due to licensing constraints, and the recommendation is pursue building a native MOOSE Delaunay triangulator routine with full functionality. Finally, a series of verification problems were performed with NEAMS physics tools. All developed capabilities will be available in the new open-source “Reactor” module of the MOOSE framework, which is accessible to any MOOSE-based NEAMS physics tool.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1995

Robotic Planetary Science Missions Enabled with Small NTR Engine/Stage Technologies

The high specific impulse (Isp) and engine thrust-to-weight ratio of liquid hydrogen (LH2)-cooled nuclear thermal rocket (NTR) engines makes them ideal for upper stage applications to difficult robotic planetary science missions. A small 15 thousand pound force (klbf) NTR engine using a uranium-zirconium-niobium 'ternary carbide' fuel (Isp approximately 960 seconds at approximately 3025K) developed in the Commonwealth of Independent States (CIS) is examined and its use on an expendable injection stage is shown to provide major increases in payload delivered to the outer planets (Saturn, Uranus, Neptune and Pluto). Using a single 'Titan IV-class' launch vehicle, with a lift capability to low Earth orbit (LEO) of approximately 20 metric tons (t), an expendable NTR upper stage can inject two Pluto 'Fast Flyby' spacecraft (PFF/SC) plus support equipment-combined mass of approximately 508 kg--on high energy, '6.5-9.2 year' direct trajectory missions to Pluto. A conventional chemical propulsion mission would use a liquid oxygen (LOX)/LH2 'Centaur' upper stage and two solid rocket 'kick motors' to inject a single PFF/SC on the same Titan IV launch vehicle. For follow on Pluto missions, the NTR injection stage would utilize a Jupiter 'gravity assist' (JGA) maneuver to launch a LOX/liquid methane (CH4) capture stage (Isp approximately 375 seconds) and a Pluto 'orbiter' spacecraft weighing between approximately 167-312 kg. With chemical propulsion, a Pluto orbiter mission is not a viable option because c inadequate delivered mass. Using a 'standardized' NTR injection stage and the same single Titan IV launch scenario, 'direct flight' (no gravity assist) orbiter missions to Saturn, Uranus and Neptune are also enabled with transit times of 2.3, 6.6, and 12.6 years, respectively. Injected mass includes a storable, nitrogen tetroxide/monomethyl hydrazine (N2O4/MMH) capture stage (Isp approximately 330 seconds) and orbiter payloads 340 to 820% larger than that achievable using a LOX/LH2-fueled injection stage. The paper discusses NTR technology and mission characteristics, shows NTR stage and payload accommodations within the 26.2 m long Titan IV payload fairing, and discusses NTR stage performance as a function of assumed cryogenic tank technology.

Borowski, Stanley K.↗
DOE OSTI2020

National Reactor Innovation Center NRC Early Site Permit Roadmap

The National Reactor Innovation Center (NRIC) at Idaho National Laboratory (INL) in Idaho Falls, Idaho, was authorized under the Nuclear Energy Innovation Capabilities Act (Public Law 115–248) to provide private sector technology developers with resources and infrastructure for testing, demonstration, and performance assessment to accelerate deployment of new advanced reactor technology concepts. NRIC’s mission is to enable clean, affordable, reliable energy by supporting U.S. government investments in nuclear energy research, development, demonstration, and commercialization of new nuclear energy systems. It intends to provide existing facilities and other undeveloped and previously developed sites at INL to advanced reactor vendors for prototype technology testing and deployment. Some of these demonstration projects may be subject to DOE authorization, while other projects may require approval by the U.S. Nuclear Regulatory Commission (NRC) prior to construction and operation. NRIC and INL are considering whether to pursue one or more ESPs to help facilitate new reactor deployment at the INL Site. An ESP would allow early investment in licensing infrastructure in advance of a reactor design being ready to deploy. Through the ESPs, NRIC would help reduce the cost and time required to deploy demonstration reactors at the INL Site. This also will allow INL to bring its nuclear and siting experience to the ESP development process allowing for better and more efficient improvements and utilization of the Site. A preapproved ESP at INL will also remove the uncertainty of NRC siting approvals from the demonstration project proponents, who may not be well positioned to complete the analyses in a timely manner. The purpose of this NRIC ESP roadmap is to provide rigor and understanding needed for INL and DOE decision-makers to make data-informed decisions on pursuing NRC ESPs in support of future advanced reactor demonstrations at INL. The roadmap provides a structured method for addressing potential challenges and identifying options for resolution.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

National Reactor Innovation Center NRC Early Site Permit Roadmap

The National Reactor Innovation Center (NRIC) at Idaho National Laboratory (INL) in Idaho Falls, Idaho, was authorized under the Nuclear Energy Innovation Capabilities Act (Public Law 115–248) to provide private sector technology developers with resources and infrastructure for testing, demonstration, and performance assessment to accelerate deployment of new advanced reactor technology concepts. NRIC’s mission is to enable clean, affordable, reliable energy by supporting U.S. government investments in nuclear energy research, development, demonstration, and commercialization of new nuclear energy systems. It intends to provide existing facilities and other undeveloped and previously developed sites at INL to advanced reactor vendors for prototype technology testing and deployment. Some of these demonstration projects may be subject to DOE authorization, while other projects may require approval by the U.S. Nuclear Regulatory Commission (NRC) prior to construction and operation. NRIC and INL are considering whether to pursue one or more ESPs to help facilitate new reactor deployment at the INL Site. An ESP would allow early investment in licensing infrastructure in advance of a reactor design being ready to deploy. Through the ESPs, NRIC would help reduce the cost and time required to deploy demonstration reactors at the INL Site. This also will allow INL to bring its nuclear and siting experience to the ESP development process allowing for better and more efficient improvements and utilization of the Site. A preapproved ESP at INL will also remove the uncertainty of NRC siting approvals from the demonstration project proponents, who may not be well positioned to complete the analyses in a timely manner. The purpose of this NRIC ESP roadmap is to provide rigor and understanding needed for INL and DOE decision-makers to make data-informed decisions on pursuing NRC ESPs in support of future advanced reactor demonstrations at INL. The roadmap provides a structured method for addressing potential challenges and identifying options for resolution.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

First-principles investigation of cerium and neodymium diffusion in BCC chromium and vanadium via vacancy-mediated transport

Lanthanide transport plays a crucial role in the performance and longevity of metallic nuclear fuels. This study examines the diffusion behavior of Ce and Nd—two major fission products—in body-centered cubic (BCC) Cr and V, which are potential liner or coating materials for mitigating fuel-cladding chemical interactions (FCCI). Using density functional theory (DFT) calculations and self-consistent mean-field (SCMF) analysis, the vacancy-mediated diffusion coefficients are evaluated. Our findings reveal that Ce and Nd act as oversized solutes and are strongly bound to vacancies in BCC Cr and V, with diffusivities in Cr significantly lower than in V and in hexagonal closed-packed (HCP) Zr, as investigated in our previous work. The activation energies for Ce and Nd diffusion are 3.39 and 3.32 eV, respectively, in BCC Cr, and 2.56 and 2.33 eV, respectively, in BCC V. Analysis of vacancy drag and partial diffusion coefficient ratios indicates a strong tendency for lanthanide enrichment at vacancy sinks in BCC Cr, and to a lesser extent in BCC V, with this effect persisting up to the melting point in Cr and remaining substantial for Nd in V at high temperatures. Under irradiation, the increase in vacancy concentration is expected to enhance lanthanide transport, potentially accelerating interactions at liner-cladding interfaces. Although BCC Cr exhibits relatively low lanthanide diffusivities under equilibrium conditions, the expected segregation tendencies under irradiation suggest that Zr liners may be a more favorable option. Further investigations using rate theory, cluster dynamics, and phase-field modeling are required to quantitatively assess the performance of these materials in reactor environments.

36 - MATERIALS SCIENCE↗
DOE OSTI2026

Direct vitrification of HLW: evaluation of candidate source tanks

The initial objective of the present work was to review Hanford HLW compositions to identify HLW streams that are good candidates for the HLW direct feed vitrification option. Another objective was to develop and identify HLW glass compositions and glass forming additive blends for vitrification of two HLW streams identified from the review that could be treated with little or no washing or other pretreatment and would be, therefore, good direct feed candidates. The HLW glass compositions and additive blends were developed to maintain high waste loadings, high processing rates, and acceptable product glass properties. This was accomplished through a combination of crucible-scale and DM100 melter system tests.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

MSTAR2019 Code Description and User's Manual

This report describes an ideal cascade model of uranium enrichment that includes side feed and side product streams and flexible input options. It is based on the original MSTAR model developed by Ed Von Halle and implemented in a Visual Basic code. The current version allows the user to specify integer numbers of stages or the stage numbers of external flows instead of assays in those flows. The computational engine is written in FORTRAN-90 and is invoked by a user-friendly GUI written in C++. This version of MSTAR has been demonstrated to operate on Linux, Mac, and Windows platforms, and has undergone significant testing and quality analysis. A number of examples are presented to illustrate the operation of the code and guide the user. The code is extremely fast, and results are returned immediately. The input and output have been specially configured for analysis by the environmental sampling team of the International Atomic Energy Agency (IAEA). A number of additional output features are included to assist the user in visualizing computational results and downloading data to files for use in other analysis software.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

DOE-NE LWRS Integrated Program Plan - Physical Security Pathway

Domestic nuclear power is facing increased financial pressures from a variety of areas and there is pressure on these utilities to reduce their cost of operation. Currently, about 20%-30% of all on-site personnel are related to physical security. The LWRS Program recognized that R&D related to physical security could play a role in providing nuclear utilities technical and staffing efficiency options to meet their physical security commitments, but utilities often lack the technical basis or the ability to create the technical basis to realize or implement these efficiencies; towards this end, the LWRS Program created the Physical Security Pathway in September 2019. The pathway performs R&D to develop methods, tools, and technologies to optimize and modernize a nuclear power facility’s security posture. The pathway will: (1) conduct research on risk-informed techniques for physical security that account for a dynamic adversary; (2) apply advanced modeling and simulation tools to better inform physical-security scenarios and reduce uncertainties in force-on-force modeling; (3) assess benefits from proposed enhancements and novel mitigation strategies and explore changes to best practices, guides, or regulation to enable modernization; and (4) enhance and provide the technical basis for stakeholders to employ new security methods, tools, and technologies.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

How To Maintain Effective IAEA Safeguards While Reducing Facility Impact

For almost twenty years, inspectors from the International Atomic Energy Agency have been visiting the K-Area Material Storage facility at the Savannah River Site where excess plutonium has been placed under IAEA safeguards as part of the United States Volunatary Offer Agreement. During that time, the mission of K-Area Complex has expanded from storage of nuclear matieral to glovebox operations in support of the National Nuclear Security Administration’s Plutonium Disposition program. With this change in mission, it is becoming increasingly challenging to manage facility staff radiation exposure and operating activities during the IAEA’s annual Physcial Inventory Verifications (PIV). This study has identified both near-term and long-term recommendations for IAEA monitoring protocols that would allow the IAEA to maintain effective safeguards of the material while decreasing the impact to the facility. A statisitical analysis has shown that reducing the number of containers measured during an annual PIV from 12 to 4 still ensures a high level of confidence in the confidence while reducing the inspector time in the facility amd with a corresponding significant reduction in personnel radiation exposure. A modified installation pattern of Remotely Monitored Sealing Arrays (RMSAs) is also proposed utilizing only the outer rows of arrays of stored material containers in the Stack Area (SA). The modified RMSA placement will provide equivalent protection, while significantly reducing personnel radiological exposure, manpower requirements and associated costs. With advances in technology, there are additional potential options in containment/surveillance and continuity of knowledge that would allow the agency to be outside of the facility during PIV measurements and have confidence in the accuracy and authencitity of the results.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2025

A Novel Dosimetry Method for Small Animal Irradiators Using 3D-printed Mouse Phantoms and Alanine Dosimeters

Abstract Accurate dosimetry is a crucial component of small animal and preclinical irradiation studies. Various dosimetry options are available but fail to characterize complex geometries and variable energy spectra of modern x-ray irradiators accurately. These options also lack national/international standards of recognition. This paper presents a novel dosimetry system, Dosequate, which uses murine phantoms embedded with alanine dosimeters and paired with x-ray energy spectra corrections to deliver consistent and accurate dosimetry measurements. This study compares Dosequate measurements against a Precision X-RAD 320 internal ion chamber and the treatment planning system of an Xstrahl Small Animal Radiation Research Platform. Results demonstrate the accuracy and reproducibility of the Dosequate system, highlighting its potential for standardizing dosimetry in preclinical research throughout the industry. Results demonstrate the accuracy and reproducibility of the Dosequate system, highlighting its potential for standardizing dosimetry in preclinical research throughout the industry. The Dosequate dosimetry method provides a robust and standardized approach for measuring absorbed dose in small animal irradiators. Its accuracy, reproducibility, and ability to account for complex irradiation geometries make it a valuable tool for preclinical research. This system has the potential to significantly improve the intercomparability of studies across different facilities and enhance the reliability of results.

Environmental Sciences & Ecology↗
DOE OSTI2020

Panel Session 120: Roundtable: Waste Management - Energy Facilities Contractor Operating Group (EFCOG WM)

This Roundtable focused on the WM EFCOG working group as they met to discuss a variety of issues of importance to US DOE waste management operations. The purpose of the WM EFCOG is to seek out and promote the best management and operating practices, cost effective technologies and disposal options for all waste streams generated at US DOE facilities whether destined for US DOE or commercial facilities. Each site representative provided a radioactive waste management lessons learned briefing. Panelists with presentations: Savannah River Site (SRS) Solid Waste Program Update (Kerri Crawford); Idaho Clean-up Project Status (Frederick Hughes); West Valley Demonstration Project Update - EFCOG Waste Management Working Group (John Rendall); Key Project Challenges and Successes (Kalli Shupe); EFCOG WMWG Challenging Waste Subgroup Update (Tammy Monday); CH TRU Difficult Waste Stream Processing Challenges (Daniel Pancake); Argonne Building 350 Clean-up Success Using Effective Work Control (Daniel Dilday); Problematic Waste Streams on the Oak Ridge Reservation (John Wrapp)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS1999

Growing crops for space explorers on the moon, Mars, or in space

An option in the long-duration exploration of space, whether on the Moon or Mars or in a spacecraft on its way to Mars or the asteroids, is to utilize a bioregenerative life-support system in addition to the physicochemical systems that will always be necessary. Green plants can use the energy of light to remove carbon dioxide from the atmosphere and add oxygen to it while at the same time synthesizing food for the space travelers. The water that crop plants transpire can be condensed in pure form, contributing to the water purification system. An added bonus is that green plants provide a familiar environment for humans far from their home planet. The down side is that such a bioregenerative life-support system--called a controlled environment life-support system (CELSS) in this paper--must be highly complex and relatively massive to maintain a proper composition of the atmosphere while also providing food. Thus, launch costs will be high. Except for resupply and removal of nonrecycleable substances, such a system is nearly closed with respect to matter but open with respect to energy. Although a CELSS facility is small compared to the Earth's biosphere, it must be large enough to feed humans and provide a suitable atmosphere for them. A functioning CELSS can only be created with the help of today's advanced technology, especially computerized controls. Needed are energy for light, possibly from a nuclear power plant, and equipment to provide a suitable environment for plant growth, including a way to supply plants with the necessary mineral nutrients. All this constitutes the biomass production unit. There must also be food preparation facilities and a means to recycle or dispose of waste materials and there must be control equipment to keep the facility running. Humans are part of the system as well as plants and possibly animals. Human brain power will often be needed to keep the system functional in spite of the best computer-driven controls. The particulars of a CELSS facility depend strongly on where it is to be located. The presence of gravity on the Moon and Mars simplifies the design for a facility on those bodies, but a spacecraft in microgravity is a much more challenging environment. One problem is that plants, which are very sensitive to gravity, might not grow and produce food in the virtual absence of gravity. However, the experience with growing super-dwarf wheat in the Russian space station Mir, while not entirely successful because of the sterile wheat heads, was highly encouraging. The plants grew well for 123 days, producing more biomass than had been produced in space before. This was due to the high photon flux available to the plants and the careful control of substrate moisture. The sterile heads were probably due to the failure to remove the gaseous plant hormone, ethylene, from the Mir atmosphere. Since ethylene can easily be removed, it should be possible to grow wheat and other crops in microgravity with the production of viable seeds. On the ground Biosphere-2 taught us several lessons about the design and construction of a CELSS facility, but Bios-3 came much closer to achieving the goals of such a facility. Although stability was never completely reached, Bios-3 was much more stable than Biosphere-2 apparently because every effort was made to keep the system simple and to use the best technology available to maintain control. Wastes were not recycled in Bios-3 except for urine, and inedible plant materials were incinerated to restore CO2 to the atmosphere. Since much meat (about 20% of calories) was imported, closure in the Bios-3 experiments was well below 100%. But then, a practical CELSS on the Moon might also depend on regular resupply from Earth. Several important lessons have been learned from the CELSS research described in this review.

Review↗
DOE OSTI2023

State-of-the-art and review of condensation heat transfer for small modular reactor passive safety: Computational studies

The small modular reactor (SMR) is a promising option with added safety features, economical manufacturing, reliable parts, portability, and scalable energy capacity that emits no greenhouse gas during its operating lifespan. The SMR safety systems, however, need to be evaluated for design and licensing. Thus, they require the verification and validation of the physics models and correlations. This study focuses on state-of-the-art condensation heat transfer analysis and a review of previous studies related to the passive containment cooling system (PCCS) of a SMR. In the PCCS of a SMR, due to its smaller size containment, filmwise condensation is dominant and therefore emphasized in this study. Furthermore, previous condensation heat transfer studies for PCCSs did not make the SMR the primary focus, so a critical review for formulating the state-of-the-art is necessary. A previous review covered experimental condensation heat transfer studies with a brief overview of associated test facilities and empirical correlations. This review covers the empirical, resistance-layer and theoretical (numerical and commercial CFD) approaches.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Study of a Proposed fissionTPC Measurement of 233 U(n,f)/ 235 U(n,f) Cross Section Ratio

A review of the available 233 U( n, f )/ 235 U( n, f ) cross section ratio data compared to the ENDF evaluation in the fast neutron energy range, 0.1–20 MeV, shows that discrepancies exist between some of the various measurements and the evaluation. The discrepancies vary and a function of incident neutron energy and range between 1–5%. In this document we review two potential options for performing new measurements of the cross section ratio to potentially resolve these discrepancies. The two options are compared and contrasted in terms of measurement accuracy and cost. The first option would utilize the fissionTPC, which was designed specifically to achieve measurement uncertainties of better than 1% to resolve a similar data–evaluation discrepancy in the 239 Pu( n, f ) cross section. While this would seem to be an ideal option in this case also, there are challenges. The fissionTPC is a complex instrument and the cost of maintenance and operations are significant. Furthermore, some of the fissionTPC equipment is nearing the end of its useful life which presents a risk to completion of a measurement campaign. The second option is to use a dual-fission chamber, which is the more traditional approach to these types of measurements. The cost is estimated to be about half that of a fissionTPC measurement, while the much simpler design of the detector reduces overall risk. The trade-off is the measurement uncertainty, which would be on order of 1–3% with a careful measurement and analysis. This uncertainty however may not be good enough to fully resolve the discrepancies. The objective of a new measurement would be the improvement in the ENDF evaluation with the end goal of improving the accuracy of models and simulations at the end of the nuclear data pipeline. An understanding of the sensitivity of these models to the data uncertainties would help guide the choice of measurement techniques.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
NASA NTRS1998

Nuclear modules for space electric propulsion

Analysis of interplanetary cargo and piloted missions requires calculations of the performances and masses of subsystems to be integrated in a final design. In a preliminary and scoping stage the designer needs to evaluate options iteratively by using fast computer simulations. The Oak Ridge National Laboratory (ORNL) has been involved in the development of models and calculational procedures for the analysis (neutronic and thermal hydraulic) of power sources for nuclear electric propulsion. The nuclear modules will be integrated into the whole simulation of the nuclear electric propulsion system. The vehicles use either a Brayton direct-conversion cycle, using the heated helium from a NERVA-type reactor, or a potassium Rankine cycle, with the working fluid heated on the secondary side of a heat exchanger and lithium on the primary side coming from a fast reactor. Given a set of input conditions, the codes calculate composition. dimensions, volumes, and masses of the core, reflector, control system, pressure vessel, neutron and gamma shields, as well as the thermal hydraulic conditions of the coolant, clad and fuel. Input conditions are power, core life, pressure and temperature of the coolant at the inlet of the core, either the temperature of the coolant at the outlet of the core or the coolant mass flow and the fluences and integrated doses at the cargo area. Using state-of-the-art neutron cross sections and transport codes, a database was created for the neutronic performance of both reactor designs. The free parameters of the models are the moderator/fuel mass ratio for the NERVA reactor and the enrichment and the pitch of the lattice for the fast reactor. Reactivity and energy balance equations are simultaneously solved to find the reactor design. Thermalhydraulic conditions are calculated by solving the one-dimensional versions of the equations of conservation of mass, energy, and momentum with compressible flow.

Difilippo, F. C.↗
DOE OSTI2020

Detector Performances Studies at Muon Collider

The Muon Collider is a possible option for the next generation of high energy collider machines. It would permit to achieve the energy frontier in lepton collisions, without occurring in significant synchrotron radiation losses as in electrons rings.Among the technological challenges in the realization of such a machine, the treatment of the beam-induced background is one of the most critical issues for the detectors.Beams with intensity of the order of $10^{12}$ muons per bunch are necessary to obtain the desired luminosity, therefore the muons decay rate is very high. Beam decay products and subsequent particles from secondary interactions with the machine elements can reach the interaction point, limiting the physical performance of the detector. A study of the beam-induced background is presented together with possible strategies for its reduction. Preliminary results on the reconstruction of a benchmark process, $\mu^+\mu^-\to H\nu\bar{\nu}\to b \bar{b}\nu\bar{\nu}$, including the beam-induced background are illustrated as demonstration of physics measurements feasibility in this harsh environment.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗