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At least 1,495 records · Page 83
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Molten Salt Reactor Fuel Depletion Tools in SCALE

Molten salt reactor depletion modeling and simulation tools have been implemented into the SCALE suite for reactor analysis and design developed by Oak Ridge National Laboratory. For depletion with material feeds and removals, additional removal and feed rate functionalities are being developed in ORIGEN to account for these flows during depletion. To track separate fuel and waste materials, additional functions and an iterative scheme are being added to the SCALE/TRITON lattice physics tool. The functionality of these tools has been demonstrated in part for previous analyses using external scripts that manage SCALE/TRITON calculations. Direct implementation of these functionalities into SCALE provides for a more ecient, more accurate material accountability methodology that is accessible to external SCALE users with a new input definition. These implementations are being tested for accuracy and to demonstrate their applicability to predefined molten salt reactor use case problems. This work demonstrates the capability of the SCALE modeling and simulation tools for reactor physics and fuel cycle analysis of molten salt reactor concepts.

Betzler, Benjamin↗
DOE OSTI2019

Verification of the ENDF/B-VII.1 and VIII.0 AMPX 1597-Group Libraries for Advanced Reactor Analysis

The SCALE/AMPX multigroup (MG) cross section processing procedure has been updated to minimize reactivity differences for various advanced thermal and fast reactor designs, as observed for the current MG libraries, resulting in excellent agreement between the calculations with the new MG libraries and the continuous-energy reference calculations. The SCALE MG calculations are widely applied to thermal spectrum light-water reactor systems, as well as fast spectrum metallic systems. Due to growing interest from industry and regulators in applying SCALE for the design of fast spectrum reactors—both sodium and molten salt—it was desirable to review and improve the SCALE/AMPX procedure for unresolved resonance self-shielded data and high-energy neutron spectra. The data were improved by generating MG unresolved resonance data based on the analytic probability table method with the narrow resonance approximation and by using very fine and intermediate group structures that are typical for fast system analysis. This study focused on verifying the improved SCALE/AMPX MG cross section processing procedure and the new AMPX 1597-group library with the ENDF/B-VII.1 and VIII.0 evaluated nuclear data. The verification was made by performing reaction rate analysis and benchmark calculations for various thermal and fast reactor systems. Results indicate that the improved SCALE/AMPX MG cross section processing and libraries provide excellent results for advanced reactor analysis.

Kim, Kang Seog↗
DOE OSTI2019

NOBLE METAL MASS TRANSPORT MODEL FOR MOLTEN SALT REACTOR ANALYSIS IN VERA-CS

Molten Salt Reactor designs require robust multiphysics modelling capabilities to further the development of this next generation technology. Due to the unique liquid fuel design, traditional reactor physics analysis codes do not capture all of the important physics of these reactors. This work builds upon a general species transport model implemented into the multiphysics core simulator suite VERA-CS, by adding source and sink models for non-soluble fission products, specifically the noble metals. Noble metal fission products are born in the fuel-salt, but do not form stable fluorides, and instead plate out on various surfaces in the reactor fuel loop. A concentration gradient diffusion-convection based mass transfer model is implemented into the general species transport model within CTF-the subchannel thermal hydraulics code in VERA-CS-and verified. Two noble metal decay chains are analyzed, and their steady state bulk and surface concentrations are calculated for a simple flow loop in CTF that is representative of the Oak Ridge National Laboratory Molten Salt Reactor Experiment (MSRE). Non-soluble species 99Mo to 99Tc captures wall plate out of decaying species, and non-soluble 135Te to soluble 135I and gaseous 135Xe captures wall loss of decaying species. Results are compared with MSRE data and the possible multiphysics effects of this phenomena are identified. Future work to enhance the robustness of the physical model and couple it with neutronics to evaluate these effects of Noble Metal mass transport is discussed.

Walker, Samuel↗
DOE OSTI2020

Retractable Sensor for Reactor Experiments

This presentation is to orient a mechanical or nuclear engineering senior design team so that they can help solve an instrumentation problem found in high power test reactors. Test reactors such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory are used to irradiate nuclear fuels and materials to evaluate performance after high levels of exposure to a reactor in-pile environment. The purpose of the experiments is to determine property changes as the materials or fuels are bombarded with fast neutrons (and thermal (slow) neutrons as well). Typically, the irradiation must take place at a very specific temperature. Sometimes other parameters are monitored as well as properties such as creep, or gas composition, etc. However, the fast neutrons cause changes in not only the materials, but also in the transducers that are placed in the neutron flux, e.g., thermocouples or optical fibers. This task will be limited to considering temperature measurements. Thermocouples experience decalibration from not only the neutron flux, but also from the very high temperatures that are sometimes measured. Optical fibers darken in a neutron or gamma flux. However, it takes quite a few hours, or days for these changes to manifest. High power test reactors typically run at a constant power and so the temperature in an experiment is fairly stable over time. Because the changes are typically very slow, even a single temperature measurement per day, would provide 95% of a perfect data set. The basic concept of this effort is to push a very small diameter thermocouple or optical fiber into the location to be measured, leave it for 30 seconds or so for it to come to equilibrium, and transmit the signal, and then pull it up and away from the high neutron flux and high temperature region. The distance the thermocouple or fiber would need to move is on the order of 50 – 100 cm. By doing this, the thermocouple junction or optical fiber would spend only a few hours in the high flux/high temperature environment over the life of the irradiation.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2021

Solar thermochemical reactor and methods of manufacture and use thereof

Disclosed herein is a solar reactor comprising a reactor member; an aperture for receiving solar radiation, the aperture being disposed in a plane on a wall of the reactor member, where the plane is oriented at any angle other than parallel relative to the centerline of the reactor member; a plurality of absorber tubes, wherein the absorber tubes are oriented such that their respective centerlines are at an angle other than 90° relative to the centerline of the reactor member; and wherein the aperture has a hydraulic diameter that is from 0.2 to 4 times a hydraulic diameter of at least one absorber tube in the plurality of absorber tubes; and a reactive material, the reactive material being disposed in the plurality of absorber tubes.

Klausner, James F.↗
DOE OSTI2021

Development and Performance Simulations of the MPACT-AGREE Code Coupling Interface for MAGNOX Reactors

MPACT is a whole-core 3D neutron transport code jointly developed by Oak Ridge National Laboratory and the University of Michigan to perform high-fidelity light water reactor (LWR) analysis. Its capability was recently extended for gas-cooled reactor analysis by coupling to the thermo-fluids code AGREE for thermal feedback. This paper presents the MPACT-AGREE code coupling for the simulation of MAGNOX-type gas-cooled graphite-moderated reactors. A coupling interface has been developed, along with methods to accurately simulate MAGNOX type reactors. The coupling mechanics were tested using two problems derived from the Calder Hall reactor. Simulations of these models demonstrated that the codes have been successfully coupled and can provide reasonable results in a tractable simulation time.

Gentry, Cole↗
DOE OSTI2022

Mobile heat pipe cooled fast reactor system

A mobile heat pipe cooled fast nuclear reactor may be configured for transportation to remote locations and may be able to provide 0.5 to 2 megawatts of power. The mobile heat pipe cooled fast reactor may contain a plurality of heat pipes that are proximate to a plurality of fuel pins inside the reactor. The plurality of heat pipes may extend out of the reactor. The reactor may be configured to be placed in a standard shipping container, and may further be configured to be contained within a cask and attached to a skid for easier transportation.

McClure, Patrick Ray↗
DOE OSTI2023

Retractable Sensors for In-Core Use in Material Test Reactors - conf paper

Material Test Reactors (MTRs) such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) are used to irradiate nuclear fuels and materials to evaluate their performance after high levels of exposure to a reactor in-core environment. The most critical tests are equipped with instrumentation leads, which allow real-time data collection. However, because of the very harsh environment inside high-power MTR experiments, there are very few sensors that can survive and maintain their calibrated readings for the time periods required to obtain the high neutron doses needed for new fuels and materials qualification. As a result, sometimes sponsoring programs are forced to accept low reliability of sensors, collecting useful data for only part of the experiment duration. The work described herein is based on the observation that MTRs normally run at constant power and the corresponding conditions within reactor experiments typically evolve relatively slowly. Therefore, even one or two measurements per day would provide a complete and representative data set. With this in mind, INL has embarked on a program to develop a mechanism capable of pushing a very small-diameter sensor (typically a thermocouple or optical fiber) into the location to be measured, leave the sensor for roughly 60 seconds to allow it to reach equilibrium and transmit the signal, then pull it up and away from the high neutron flux and high-temperature region. Small-diameter capillary tubes, up to 8 m long, are used to guide the sensors to the appropriate locations. These capillary tubes serve as essentially very deep, thin-walled thermowells. The distance a thermocouple or optical fiber would need to traverse is on the order of 40 - 80 cm. By adopting this infrequent cycling strategy, the thermocouple or optical fiber would spend only a few hours in the high-neutron flux/high-temperature environment over the duration of even the longest irradiation experiment. To date, INL has developed two styles of drive mechanisms. The first is based on friction drive wheels which drive the sensors in a manner similar to a small MIG welder. This has the advantage of being able to accommodate a very long insertion length. The second is based on a ball screw drive and has the advantages of positive attachment and being able to move more than one sensor at a time. Both drive mechanisms have been fabricated and tested in a laboratory setting. Both systems can handle hard mineral insulated cable (such as thermocouples) or optical fibers encased in small diameter tube. The sizes tested to date are 1 - 1.6 mm diameter. Work in this area is ongoing with an eye toward demonstration in the Massachusetts Institute of Technology's MITR reactor, followed by deployment in an ATR irradiation experiment.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2023

Retractable Sensors for In-Core Service in Material Test Reactors

Material Test Reactors (MTRs) such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) are used to irradiate nuclear fuels and materials to evaluate their performance after high levels of exposure to a reactor in-core environment. The most critical tests are equipped with instrumentation leads, which allow real-time data collection. However, because of the very harsh environment inside high-power MTR experiments, there are very few sensors that can survive and maintain their calibrated readings for the time periods required to obtain the high neutron doses needed for new fuels and materials qualification. As a result, sometimes sponsoring programs are forced to accept low reliability of sensors, collecting useful data for only part of the experiment duration. The work described herein is based on the observation that MTRs normally run at constant power and the corresponding conditions within reactor experiments typically evolve relatively slowly. Therefore, even one or two measurements per day would provide a complete and representative data set. With this in mind, INL has embarked on a program to develop a mechanism capable of pushing a very small-diameter sensor (typically a thermocouple or optical fiber) into the location to be measured, leave the sensor for roughly 60 seconds to allow it to reach equilibrium and transmit the signal, then pull it up and away from the high neutron flux and high-temperature region. Small-diameter capillary tubes, up to 8 m long, are used to guide the sensors to the appropriate locations. These capillary tubes serve as essentially very deep, thin-walled thermowells. The distance a thermocouple or optical fiber would need to traverse is on the order of 40 - 80 cm. By adopting this infrequent cycling strategy, the thermocouple or optical fiber would spend only a few hours in the high-neutron flux/high-temperature environment over the duration of even the longest irradiation experiment. To date, INL has developed two styles of drive mechanisms. The first is based on counter-rotating wheels which drive the sensors in a manner similar to a small MIG welder. This has the advantage of being able to accommodate a very long insertion length. The second is based on a ball screw drive and has the advantages of positive attachment and being able to move more than one sensor at a time. Both drive mechanisms have been fabricated and tested in a laboratory setting. Both systems can handle hard mineral insulated cable (such as thermocouples) or optical fibers encased in small diameter tube. The sizes tested to date are 1 - 1.6 mm diameter. Work in this area is ongoing with an eye toward demonstration in the Massachusetts Institute of Technology's MITR reactor, followed by deployment in an ATR irradiation experiment.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2023

Feasibility of Power Ramp Testing in the Advanced Test Reactor

For decades, the Halden Boiling Water Reactor (HBWR) in Norway was an international resource for assessing nuclear fuels and materials behavior and its unexpected shutdown in 2018 represented a significant loss in experimental capability for prototypical irradiation testing. In the aftermath of the closure, a study was performed to assess capability gaps related to the Accident Tolerant Fuels (ATF) program. It was concluded that the primary capability gaps left by the closure of the HBWR were the loss of prototypic in-pile Light Water Reactor (LWR) loops that provide operational transient testing and in-pile Loss-of-Coolant Accident (LOCA) testing capabilities. The study also concluded that the Advanced Test Reactor (ATR) and Transient Reactor Test Facility (TREAT) likely have the necessary key capabilities to absorb the breadth of the HBWR mission gaps related to the ATF program, but additional investments in experimental infrastructure were needed. One such investment is the design and installation of additional pressurized water loops in the medium-I positions of the ATR, hereafter referred to as I-Loops, to support power ramp testing and testing in Boiling Water Reactor (BWR) conditions. This paper gives an overview of the planned ATR I-Loops and assesses the feasibility of performing power ramp testing inside such a loop.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Retractable Sensors for In-Core Service in Material Test Reactors

Material Test Reactors (MTRs) such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) are used to irradiate nuclear fuels and materials to evaluate their performance after high levels of exposure to a reactor in-core environment. The most critical tests are equipped with instrumentation leads, which allow real-time data collection. However, because of the very harsh environment inside high-power MTR experiments, there are very few sensors that can survive and maintain their calibrated readings for the time periods required to obtain the high neutron doses needed for new fuels and materials qualification. As a result, sometimes sponsoring programs are forced to accept low reliability of sensors, collecting useful data for only part of the experiment duration. The work described herein is based on the observation that MTRs normally run at constant power and the corresponding conditions within reactor experiments typically evolve relatively slowly. Therefore, even one or two measurements per day would provide a complete and representative data set. With this in mind, INL has embarked on a program to develop a mechanism capable of pushing a very small-diameter sensor (typically a thermocouple or optical fiber) into the location to be measured, leave the sensor for roughly 60 seconds to allow it to reach equilibrium and transmit the signal, then pull it up and away from the high neutron flux and high-temperature region. Small-diameter capillary tubes, up to 8 m long, are used to guide the sensors to the appropriate locations. These capillary tubes serve as essentially very deep, thin-walled thermowells. The distance a thermocouple or optical fiber would need to traverse is on the order of 40 – 80 cm. By adopting this infrequent cycling strategy, the thermocouple or optical fiber would spend only a few hours in the high-neutron flux/high-temperature environment over the duration of even the longest irradiation experiment. To date, INL has developed two styles of drive mechanisms. The first is based on counter-rotating wheels which drive the sensors in a manner similar to a small MIG welder. This has the advantage of being able to accommodate a very long insertion length. The second is based on a ball screw drive and has the advantages of positive attachment and being able to move more than one sensor at a time. Both drive mechanisms have been fabricated and tested in a laboratory setting. Both systems can handle hard mineral insulated cable (such as thermocouples) or optical fibers encased in small diameter tube. The sizes tested to date are 1 – 1.6 mm diameter. Work in this area is ongoing with an eye toward demonstration in the Massachusetts Institute of Technology’s MITR reactor, followed by deployment in an ATR irradiation experiment.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2016

CyBORG: An ORIGEN-Based Reactor Analysis Capability for Cyclus

One of the key weaknesses of many current nuclear fuel cycle simulation tools is in the inability to treat reactor-based materials feedbacks within fuel cycle systems. Rather, many of these tools rely on user-provided 'recipes' to describe the fission and transmutation of materials within a reactor. While sufficient for steady-state simulations (such as fuel cycle options at equilibrium), recipe-based models fall short when attempting to evaluate fuel cycle systems under transition. For example, within the context of the Fuel Cycle Options Evaluation and Screening Study, fuel cycle options such as EG29 (fast critical systems with a conversion ratio greater than unity which would drive thermal critical systems with recovered U and Pu) require physics-based treatments of materials within reactors in order to capture the full dynamics of material transformations within such actinide-based recycle scenarios. Similarly, other physics-based questions include system perturbations such as changing average fleet discharge burnup as well as accurate evaluations of available fuel materials to support fuel cycle transitions. In order to provide fuel cycle analysis the ability to evaluate these types of scenarios, a new ORIGEN-based reactor module for the Cyclus nuclear fuel cycle simulator is being developed known as CyBORG (Cyclus-Based ORiGen). CyBORG leverages recent modernization efforts in the Origen API to allow users to directly employ ORIGEN's validated depletion capabilities and extensive nuclear fuel data libraries in order to perform dynamic, physics-based reactor transformations to fuel materials within a fuel cycle simulation context. This paper will provide a summary of recent developments in this effort, including its present and proposed capabilities and overall design strategy. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Pressurized-Water Reactor Core Design using Multi-Objective Plant Fuel Reload Optimization Platform

The United States (U.S.) Department of Energy (DOE) Light Water Reactor Sustainability (LWRS) Program Risk-Informed Systems Analysis (RISA) Pathway Plant Reload Optimization Project aims to develop an integrated, comprehensive framework offering an all-in-one solution for reload evaluations with a special focus on optimization of core design. The optimization of the fuel loading pattern is one of the most important considerations in reducing the amount of new fuel used in the core. Due to thousands of possible options of core configuration, finding optimal solutions is an unachievable task for a human. The Plant ReLoad Optimization (PRLO) platform which supports artificial-intelligence-based reactor core designing is now fully capable of handling realistic problems. The PRLO Platform development project aims to build a reactor core design tool that includes reactor safety and fuel performance analyses and uses artificial intelligence to support the optimization of core design solutions. The NSGA-II (Non-dominated Sorting Genetic Algorithm-II) optimizer was developed and tested within RAVEN (Risk Analysis and Virtual Environment) to handle many constraints by using an augmented objectives methodology. The demonstration was performed with constrained multi-objective optimization of a 17 × 17 pressurized-water reactor core loading patterns to minimize fuel cost and maximize fuel cycle length.

42 ENGINEERING↗
DOE OSTI2024

National Reactor Innovation Center Annual Report

The National Reactor Innovation Center (NRIC), established in August 2019, is a national United States (U.S.) Department of Energy (DOE) program. NRIC’s mission is to partner with industry and national laboratories to bridge the gap between the concept, demonstration, and commercialization of advanced nuclear technology. NRIC accomplishes this through building or enhancing existing DOE infrastructure to support the testing of components and systems that are key to successfully deploying advanced nuclear technology. NRIC works to inspire stakeholders and the public, empower innovators, and deliver successful outcomes through efficient collaboration and coordination with partners. NRIC’s vision is that by 2028, NRIC will be partnered with industry and accelerating the demonstration and deployment of advanced nuclear technology using DOE national laboratory infrastructure and expertise. NRIC will establish four new experimental facilities and two large reactor test beds for integrated technology demonstrations and experimentation by 2028 and complete two advanced nuclear technology tests by 2030. Achieving this vision will enable urgently needed abundant and affordable clean energy both domestically and internationally. NRIC’s success will inspire our nation and the global community to embrace the promising contribution of innovative nuclear reactor technologies to the clean energy economy and re-establish the U.S. as the global leader in advanced nuclear energy. NRIC is tasked with expediting the development of advanced nuclear energy technologies by bringing together private-sector technology developers and the world-class capabilities of the DOE national laboratory system. Through this program, the U.S. private sector is given access to the physical infrastructure available at DOE national laboratories to test and demonstrate their reactor concepts. NRIC works closely with the Gateway for Accelerated Innovation in Nuclear (GAIN),; which is the DOE-Nuclear Energy (NE) program that grantings access to technical, regulatory, and financial support for commercializing nuclear energy. As observed in Figure 1, NRIC builds upon these new reactor concepts and technology successes to effectively strengthen U.S. nuclear leadership.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2025

Employing a Hardware-in-the-Loop Approach to Realize a Fully Homomorphic Controller for a Small Modular Advanced High Temperature Reactor

This paper addresses the cybersecurity challenges of advanced nuclear reactors by integrating fully homomorphic encryption (FHE) into their control systems, enabling encrypted processing of control signals without compromising functionality. Advanced nuclear reactors, including Small Modular Reactors (SMRs) and microreactors, aim to achieve autonomous and remote operations, reducing costs and enhancing competitiveness. However, these advancements expand the attack surface for cyberattacks, particularly in autonomous and remote operation scenarios. Cyberattacks can exploit vulnerabilities to manipulate physical processes, causing shutdowns, asset damage, or public harm. Such attacks begin with passive reconnaissance, where adversaries intercept communications or observe behaviors to gather information, which is then leveraged to execute cyber-physical attacks by injecting malicious commands. Nuclear power must adopt cybersecurity protection measures to secure the integrity and availability of their digital control systems. This paper demonstrates the application of FHE to secure operations by enabling encrypted processing of sensitive signals and parameters -- ensuring privacy without exposing data. FHE supports secure mathematical operations on encrypted data without requiring decryption. Using a hardware-in-the-loop (HIL) approach, this paper implements an FHE-integrated controller on a BeagleBone Black (BBB) controlling a simulation of the Small Modular Advanced High Temperature Reactor (SmAHTR). By doing so, the encrypted controller protects the integrity of critical set points and control signals during transmission and processing. Thus, FHE-integrated controllers enhance secure operations of advanced nuclear reactors while maintaining functionality.

control systems↗
DOE OSTI2025

Development for Integrated System-Level Analysis Capabilities in SAM for Molten Salt Reactors

In recent years, there has been renewed interest in Molten Salt Reactors (MSRs) for their potential advantages compared to reactors that rely on solid fuel. In response to such interest, many methods and codes have been developed to capture the unique features of MSRs. Among them, the System Analysis Module (SAM) is a modern system analysis tool that provides fast-running, modest-fidelity, whole-plant transient analysis capabilities, essential for fast-turnaround design scoping and engineering analyses of advanced reactor concepts. For liquid-fuel MSRs, the complex physics and chemistry involved in MSR operation—such as reactor kinetics, fluid flow, heat transfer, and salt composition dynamics—pose significant challenges for system-level modeling. Specific modeling capabilities are needed for system-level transient simulation. This paper presents recent advancements in SAM capability enhancements for system-level modeling of MSRs, focusing on improved simulation fidelity, computational efficiency, and multi-physics integration. Key enhancements include the development of species transport, Delayed Neutron Precursor (DNP) drift, modified Point Kinetics Equations (PKE), decay heat modeling, key fission product behavior, salt corrosion, and thermal-hydraulic coupling, as well as code robustness and performance enhancements for MSR applications. The code enhancement allows for better predictive accuracy in safety analysis, transient behavior, and operational optimization, thus supporting the design and licensing of next-generation MSRs. Results from case studies are presented to demonstrate the benefits of these enhancements in accurately capturing key reactor transient behaviors.

Hu, Rui (ORCID:0000000237712920)↗
DOE OSTI2025

Restructuring and Optimizing Reactor Building Lifting Processes & Designing a Solution for an Overhead Door Handling Forklift Attachment

This poster presents two innovative projects aimed at enhancing operational efficiency and safety at the Advanced Test Reactor (ATR) Complex. The first project focuses on restructuring the existing lift book used for hoisting operations within the ATR Reactor Building. The current lift book, a cumbersome 150+ page document, is being transformed into a more efficient format using charts that allow for quick identification of maximum lift heights based on object footprint and weight. This new method also considers various parameters such as floor integrity and reactor status, dividing entries into four distinct charts to improve usability and safety during lifts. The second project involves designing a specialized forklift attachment for handling an overhead door at the ATR Reactor Building. The attachment is engineered to securely lift and lower a 1200lb, 14ft long door, ensuring safe replacement operations. The design process included research on forklift attachment codes, modeling in Autodesk Inventor, and testing through Finite Element Analysis (FEA) and hand calculations to confirm the attachment's structural integrity. Future work includes completing detailed drawings and an Engineering Calculations Analysis and Review (ECAR) document to proceed with manufacturing and assembly. Together, these projects demonstrate a commitment to optimizing reactor building operations through innovative engineering solutions and safety considerations.

42 - ENGINEERING↗
DOE OSTI2015

High Temperature Gas-Cooled Test Reactor Options Status Report

Preliminary scoping calculations are being performed for a 100 MWt gas-cooled test reactor. The initial design uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to identify some reactor design features to investigate further. Current status of the effort is described.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗