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DOE OSTI2024

Refining Absorber Shroud Geometry to Maximize Power Output and Reduce Power Peaking in ATF Test Train

In the wake of the Fukushima Daiichi nuclear power plant accident in 2011, the accident tolerant fuels (ATF) program was initiated to enhance the safety of light-water reactor fuels, placing significant emphasis on cladding. A crucial step for the broad implementation of ATF in commercial reactors involves irradiation testing of the fuel designs. ATF-2D, the latest experiment in the ATF series, is slated to undergo irradiation in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). ATF-2D is a joint effort of the INL with industry partners General Electric Global Research; Framatome; General Atomics; the Japan Atomic Energy Agency; Hitachi-GE Nuclear Energy, Ltd; Global Nuclear Fuel-Japan Co., Ltd; Nippon Nuclear Fuel Development Co., Ltd; and Mitsubishi Heavy Industries, Ltd. The ATF-2D test train design consists of four tiers, each housing six rodlets. The device will be inserted in Loop 2A within the central flux trap of the ATR, and is anticipated to undergo irradiation throughout three 60-day cycles. Typical pressurized water reactor conditions will be emulated during the irradiation. The objective of the work presented here is to dimension neutron-absorbing hafnium (Hf) components surrounding the fuel rodlets, such that the axial power profile is flattened, while simultaneously ensuring that the total fission power output of the entire test train remains below 200 kW.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Refining Absorber Shroud Geometry to Maximize Power Output and Reduce Power Peaking in ATF Test Train

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2018

Fuel performance of chromium-coated zirconium alloy and silicon carbide accident tolerant fuel claddings

The U.S. Department of Energy’s Accident Tolerant Fuel program is focused on extending the time for fuel failure during postulated severe accidents compared to the standard UO2-Zr alloy fuel system. This research investigates the feasibility of two different chromium-coated cladding concepts, one of which is zirconium-alloy based and the other is composite-SiC based. Both claddings had 50µm coatings, deducted from the base layer thicknesses. The claddings were reported on, using the multi- physics fuel performance tool BISON, under steady-state PWR operating conditions as well as two transients: power ramp and a loss-of-coolant accident (LOCA). The chromium-coated claddings showed comparable thermo-mechanical performance to the reference Zircaloy-4 cladding. As chromium is reported to provide an order of magnitude improvement in oxidation resistance, it is expected to be a better alternative in accident scenarios. Simulation results for both concepts call for further experimental investigation as well as modeling of beyond design-basis accidents.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2019

Materials-to-device design of hybrid metal-polymer heat exchanger tubes for low temperature waste heat recovery

Heat exchangers for low temperature (<150 °C) waste heat recovery (WHR) are challenging due to severe constraints on materials costs, manufacturing methods, and maintenance opportunities after installation. Here, the use of polymers instead of metals, offers an avenue for cost reduction in WHR applications, provided the typical low thermal conductivity of polymers (~0.2 W m –1 K –1 ) that results in poor overall heat transfer coefficient can be overcome. Previous approaches to enhance the thermal conductivity of polymers remain problematic in terms of cost, thermal, and thermomechanical considerations. In this work, we propose a novel hybrid metal-polymer composite assembled from strips of polymer and copper. Polymer strips with metal-clad edges are wound helically and the interfaces between the strips joined to obtain a tube for a cross-flow heat exchanger. Interfaces are designed to enhance effective heat conduction across the material and increase the overall heat transfer coefficient. Through numerical simulations, we obtain metal-polymer layouts that achieve wall effective thermal conductivity ~1 W m –1 K –1 at ~23% or ~35% volume fraction of copper or aluminum, respectively. The conductivity enhancement is sufficient for typical low temperature WHR, resulting in a ~20% enhancement in the overall heat transfer coefficient compared to a similar all-polymer heat exchanger. Thermomechanical simulations revealed that the composite pipe designs can withstand >1.4 MPa internal pressure, under the normal operating conditions of the heat exchanger. Our work provides guidelines for designing macroscopic metal-polymer composites that can be adapted for specific requirements such as reduced materials costs, ease of manufacturing, and heat transfer enhancement. In conclusion, the optimized design of hybrid metal-polymer heat exchanger tubes potentially provides a scalable and cost-effective route toward harvesting waste heat from low temperature sources.

42 ENGINEERING↗

DOE OSTI2016

Fuel rod thermo-mechanic analysis of Kuosheng NPP in MSIVC transient with RELAP5 MOD3.3/Fraptran/Python code in SNAP interface

Kuosheng NPP (Nuclear Power Plant) is located on the northern coast of Taiwan. Its nuclear steam supply system (NSSS) is a type of boiling water reactor (BWR/6) designed and built by General Electric on a twin unit concept. Each unit includes two loops of recirculation piping and four main steam lines, with the thermal rated power of 2894 MWt. Unit 1 started SPU (Stretch Power Uprate) from Cycle 24 and Unit 2 started SPU from Cycle 23. The operating power is 104.7% of the OLTP (Original Licensed Thermal Power), which is 3030 MWt now. In order to estimate the safety of Kuosheng NPP, the methodology of Kuosheng NPP SPU safety analysis model was developed. This methodology is shown in Fig. 1. There are three main steps considered in this methodology. The first step is the development of the Kuosheng NPP SPU RELAP5 model. The next step is the main steam line isolation valve closure (MSIVC) analysis of Kuosheng NPP SPU RELAP5 model in order to confirm the maximum pressure of vessel below the acceptance limit of 9.58 MPa. The final step is the fuel rods integrity analysis of FRAPTRAN under the above conditions. From FRAPTRAN analysis results, we find that the peak value of cladding hoop strain is smaller than 0.01, the acceptance limit; furthermore, the fuel enthalpy values are smaller than 150 cal/g. These two results indicate that the fuel rods are still in safe condition after turbine trip accident. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Status of Eddy Current Developments at INL

Both standard and advanced Light Water Reactor (LWR) fuel systems are composed of fuel pellets enclosed in a metallic cladding. The fuel cladding represents the first containment of the fission products and the first safety barrier. Upon progression of irradiation, the cladding properties deteriorate, mainly due to the pick-up of hydrogen, which causes embrittlement, and the consequence corrosion that reduced the metallic substrate. Those changes in the properties combined with the harsh environment, e.g., fatigue and fretting occurring during operation, might eventually lead to loss of hermeticity. Therefore, it is important to be able to quantify the amount of corrosion taking place and to detect and size structural discontinuity that can promote cladding failure. Eddy current testing is one of the non-destructive methods employed for this purpose. Inspections are carried out using special probes, which are designed and optimized to inspect a tube with well-defined geometrical and physical characteristics, i.e. external diameter, wall thickness, material electric conductivity and magnetic permeability. The inspection goals are mainly twofold: first, the determination of the amount of corrosion by measurement of the external oxide thickness; second, the detection of flaws in the tube wall, ether on the external or the internal surface of the tube. This report summarizes the available capabilities for eddy current testing of LWR fuel claddings, identifies needs and are of developments for testing of ATF claddings.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2014

Accident Tolerant Fuel Analysis

Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about light water reactor design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the Risk Informed Safety Margin Characterization (RISMC) Pathway research and development (R&D) is to support plant decisions for risk-informed margins management by improving economics and reliability, and sustaining safety, of current NPPs. Goals of the RISMC Pathway are twofold: (1) Develop and demonstrate a risk-assessment method coupled to safety margin quantification that can be used by NPP decision makers as part of their margin recovery strategies. (2) Create an advanced “RISMC toolkit” that enables more accurate representation of NPP safety margin. In order to carry out the R&D needed for the Pathway, the Idaho National Laboratory is performing a series of case studies that will explore methods- and tools-development issues, in addition to being of current interest in their own right. One such study is a comparative analysis of safety margins of plants using different fuel cladding types: specifically, a comparison between current-technology Zircaloy cladding and a notional “accident-tolerant” (e.g., SiC-based) cladding. The present report begins the process of applying capabilities that are still under development to the problem of assessing new fuel designs. The approach and lessons learned from this case study will be included in future Technical Basis Guides produced by the RISMC Pathway. These guides will be the mechanism for developing the specifications for RISMC tools and for defining how plant decision makers should propose and evaluate margin recovery strategies.

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2015

Preliminary Evaluation of FeCrAl Cladding for Accident Tolerant Fuel Concepts

Since the accident at the Fukushima Daiichi Nuclear Power Station, enhancing the accident tolerance of light water reactors (LWRs) has become an important research topic. In particular, the community is actively developing enhanced fuels and cladding for LWRs to improve safety in the event of accidents in the reactor or spent fuel pools. Fuels with enhanced accident tolerance are those that, in comparison with the standard UO2-zirconium alloy system, can tolerate loss of active cooling in the reactor core for a considerably longer time period during design-basis and beyond design-basis events while maintaining or improving the fuel performance during normal operations and operational transients. This report presents early work in developing thermal and mechanical models for FeCrAl, a proposed cladding material. BISON, the finite element-based fuel performance code in development at Idaho National Laboratory, was used to compare results from normal operation conditions with UO2/Zr-4 behavior. In addition, sensitivity studies are presented for evaluating the relative importance of material parameters such as elastic constants and thermal conductivity in order to provide guidance on future experiments for this material.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2012

Operational Challenges of Extended Dry Storage of Spent Nuclear Fuel - 12550

As a result of the termination of the Yucca Mountain used fuel repository program and a continuing climate of uncertainty in the national policy for nuclear fuel disposition, the likelihood has increased that extended storage, defined as more than 60 years, and subsequent transportation of used nuclear fuel after periods of extended storage may become necessary. Whether at the nation's 104 nuclear energy facilities, or at one or more consolidated interim storage facilities, the operational challenges of extended storage and transportation will depend upon the future US policy for Used Fuel Management and the future Regulatory Framework for EST, both of which should be developed with consideration of their operational impacts. Risk insights into the regulatory framework may conclude that dry storage and transportation operations should focus primarily on ensuring canister integrity. Assurance of cladding integrity may not be beneficial from an overall risk perspective. If assurance of canister integrity becomes more important, then mitigation techniques for potential canister degradation mechanisms will be the primary source of operational focus. If cladding integrity remains as an important focus, then operational challenges to assure it would require much more effort. Fundamental shifts in the approach to design a repository and optimize the back-end of the fuel cycle will need to occur in order to address the realities of the changes that have taken place over the last 30 years. Direct disposal of existing dual purpose storage and transportation casks will be essential to optimizing the back end of the fuel cycle. The federal used fuel management should focus on siting and designing a repository that meets this objective along with the development of CIS, and possibly recycling. An integrated approach to developing US policy and the regulatory framework must consider the potential operational challenges that they would create. Therefore, it should be integral to these efforts to redefine retrievability to apply to the dual purpose cask, and not to apply to individual assemblies. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

Rod internal pressure quantification and distribution analysis using Frapcon

This report documents work performed supporting the Department of Energy (DOE) Office of Nuclear Energy (NE) Fuel Cycle Technologies Used Fuel Disposition Campaign (UFDC) under work breakdown structure element 1.02.08.10, ST Analysis. In particular, this report fulfills the M4 milestone M4FT- 15OR0810036, Quantify effects of power uncertainty on fuel assembly characteristics, within work package FT-15OR081003 ST Analysis-ORNL. This research was also supported by the Consortium for Advanced Simulation of Light Water Reactors (http://www.casl.gov), an Energy Innovation Hub (http://www.energy.gov/hubs) for Modeling and Simulation of Nuclear Reactors under U.S. Department of Energy Contract No. DE-AC05-00OR22725. The discharge rod internal pressure (RIP) and cladding hoop stress (CHS) distributions are quantified for Watts Bar Nuclear Unit 1 (WBN1) fuel rods by modeling core cycle design data, operation data (including modeling significant trips and downpowers), and as-built fuel enrichments and densities of each fuel rod in FRAPCON-3.5. A methodology is developed which tracks inter-cycle assembly movements and assembly batch fabrication information to build individual FRAPCON inputs for each evaluated WBN1 fuel rod. An alternate model for the amount of helium released from the zirconium diboride (ZrB2) integral fuel burnable absorber (IFBA) layer is derived and applied to FRAPCON output data to quantify the RIP and CHS for these types of fuel rods. SCALE/Polaris is used to quantify fuel rodspecific spectral quantities and the amount of gaseous fission products produced in the fuel for use in FRAPCON inputs. Fuel rods with ZrB2 IFBA layers (i.e., IFBA rods) are determined to have RIP predictions that are elevated when compared to fuel rod without IFBA layers (i.e., standard rods) despite the fact that IFBA rods often have reduced fill pressures and annular fuel pellets. The primary contributor to elevated RIP predictions at burnups less than and greater than 30 GWd/MTU is determined to be the total fuel rod void volume and the amount of released fission gas in the fuel rod, respectively. Cumulative distribution functions (CDFs) are prepared from the distribution of RIP and CHS predictions for all standard and IFBA rods. The provided CDFs allow for the determination of the portion of WBN1 fuel rods that exceed a specified RIP or CHS limit. Results are separated into IFBA and standard rods so that the two groups may be analyzed individually. FRAPCON results are provided in sufficient detail to enable the recalculation of the RIP while considering any desired plenum gas temperature, total void volume, or total amount of gas present in the void volume. A method to predict the CHS from a determined or assumed RIP is also proposed, which is based on the approximately linear relationship between the CHS and the RIP. Finally, improvements to the computational methodology of FRAPCON are proposed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2021

Investigation of Thermal-Hydraulic Effects of Dry Storage Canister Helium Backfill Loss Using the Horizontal Dry Cask Simulator

A previous investigation produced data sets that can be used to benchmark the codes and best practices presently used to determine cladding temperatures and induced cooling air flows in modern horizontal dry storage systems. The horizontal dry cask simulator (HDCS) was designed to generate this benchmark data and add to the existing knowledge base. The objective of the previous HDCS investigation was to capture the dominant physics of a commercial dry storage system in a well-characterized test apparatus for a wide range of operational parameters. The close coupling between the thermal response of the canister system and the resulting induced cooling air flow rate was of particular importance. The previous investigation explored these parameters using helium backfill at 100 kPa and 800 kPa pressure as well as air backfill with a series of simulated decay heats. The helium tests simulated a horizontal dry cask storage system at normal storage conditions with either atmospheric or elevated backfill pressure, while the air tests simulated horizontal storage canisters following a complete loss of helium backfill, in which case the helium would be replaced by air. The present HDCS investigation adds to the previous investigation by exploring steady-state conditions at various stages of the loss of helium backfill from a horizontal dry cask storage system. This is achieved by using helium/air blends as a backfill in the HDCS and running a series of tests using various simulated decay heats to explore the effects of relative helium/air molar concentration on the thermal response of a simulated horizontal dry cask storage system. A total of twenty tests were conducted where the HDCS achieved steady state for various assembly powers, representative of decay heat. The power levels tested were 0.50, 1.00, 2.50, and 5.00 kW. All tests were run at 100 kPa vessel pressure. The backfill gases used in these tests are given in this report as a function of mole fraction of helium (He), balanced by air: 1.0, 0.9, 0.5, 0.1, and 0.0 He. Steady-state conditions (where the steady-state start condition is defined as where the change in temperature with respect to time for the majority of HDCS components is less than or equal to 0.3 K/h) were achieved for all test cases.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Atmospheric pressure ionisation - mass spectrometry for argon online monitoring in ASTRID

Safety, operability, waste transmutation and mastered investment costs are key issues for the development of generation IV reactors, especially sodium cooled fast reactors. To achieve a development scheduled around 2040, the nuclear energy division of CEA, France, is studying a prototype called ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) which is in its conceptual design phase. To achieve a high level of maturity, In Service Inspection and Repair is very important, in particular fuel cladding failure detection. If such a failure occurs, fission products Xenon and Krypton are released. Their fast and sensitive detection is a challenge. The use of a commercial API-MS is evaluated in this work. Detection limits below 10 ppt have been measured for Kr 84 and for Xe 132. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2012

The SAS4A/SASSYS-1 Safety Analysis Code System

SAS4A/SASSYS-1 is a software simulation tool used to perform deterministic safety analysis of anticipated events as well as design basis and beyond design basis accidents for advanced nuclear reactors. This software can be used to assess the safety of a prescribed reactor design, but it cannot be used to configure a design to meet targeted performance objectives. Detailed, mechanistic models of steady-state and transient thermal,hydraulic, kinetic, and mechanical phenomena are employed to describe the response of the reactor core, the reactor primary and secondary coolant loops, the reactor control and protection systems and the balance-of-plant to accidents caused by changes in coolantflow, oss of heat rejection, or reactivity insertion. The consequences of single and double-fault accidents can be modeled, including fuel and coolant heating, fuel and cladding mechanical behavior, core reactivity feedbacks, coolant loop performance including natural circulation, and decay heat removal. Analyses are typically terminated upon demonstration of reactor and plant shutdown to permanently coolable conditions or upon violation of design basis margins. The objective of the analysis is to quantify accident consequences as measured by the transient behavior of system performance parameters, such as fuel and cladding temperatures, reactivity, and cladding strain. Originally developed for analysis of sodium cooled reactors with oxide fuel clad by stainless steel, the models In SAS4A/SASSYS-1 were subsequently extended and specialized to metallic fuel clad with advanced alloys and to several other coolant options including lead and LBE. METHOD OF SOLUTION: Each subassembly in the reactor core is represented with single or multiple-pin models. One channel represents one or more fuel pins and many channels are employed for a whole-core representation. Heat transfer in each pin is modeled with a two-dimensional (r/z) heat conduction equation. Single and two-phase coolant thermal-hydraulics are simulated with a unique, one-dimensional (axial) multiple-bubble liquid metal boiling model. The transient fuel and cladding mechanical oehavior models are integrated with fission product production,release, and transport models. UNIQUE FEATURES: The physical models in SAS4AJSASSYS-1 are highly detailed 1umerical representations of reactor accident conditions based on extensive laboratory and test reactor results. The models are specialized to liquid metal (sodium) cooled fast reactors with oxide or metallic fuel clad with stainless steel.

Source record↗

DOE OSTI2020

Pitch Variation Experiments in Water-Moderated Square-Pitched U(6.90)O 2 Fuel Rod Lattices with Fuel to Water Volume Ratios Spanning 0.08 to 0.67

The US Department of Energy (DOE) Nuclear Energy Research Initiative funded the design and construction of the Seven Percent Critical Experiment (7uPCX) at Sandia National Laboratories. The start-up of the experiment facility and the execution of the experiments described here were funded by the DOE Nuclear Criticality Safety Program. The 7uPCX is designed to investigate critical systems with fuel for light water reactors in the enrichment range above 5 % 235 U. The 7uPCX assembly is a water-moderated and -reflected array of aluminum-clad square-pitched U(6.90 %)O 2 fuel rods. Other critical experiments performed in the 7uPCX assembly are documented in LEU-COMP-THERM-078, LEU-COMP-THERM-080, LEU-COMPTHERM- 096, LEU-COMP-THERM-097, and LEU-COMP-THERM-101. The fuel used in these experiments was fabricated using unirradiated 6.90 % enriched UO 2 fuel pellets from fuel elements designed to be used in the internal nuclear superheater section of the Pathfinder boiling water reactor operated in South Dakota by the Northern States Power Company in the 1960s. The fuel elements were obtained from The Pennsylvania State University where they had been stored for many years. The fuel pellets in those fuel elements were removed from the original Incoloy cladding and reclad in 3003 aluminum tubes and end caps for use in the experiments reported here. The purpose of these experiments was to measure the effects of decreasing the fuel-to-water volume ratio on the critical array size. This was accomplished by removing fuel rods from fully fueled configurations, effectively increasing the pitch of the fuel arrays in the assembly. The fuel rod pitch variations provided assembly configurations that ranged from strongly undermoderated to slightly overmoderated.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

SAS4A/SASSYS-1 VERSION 5.2 SAFETY ANALYSIS CODE SYSTEM

SAS4A/SASSYS-1 is a software simulation tool used to perform deterministic safety analysis of anticipated events as well as design basis and beyond design basis accidents for advanced nuclear reactors. This software can be used to assess the safety of a prescribed reactor design, but it cannot be used to configure a design to meet targeted performance objectives. Detailed, mechanistic models of steady-state and transient thermal, hydraulic, kinetic, and mechanical phenomena are employed to describe the response of the reactor core, the reactor primary and secondary coolant loops, the reactor control and protection systems and the balance-of-plant to accidents caused by changes in coolant flow, oss of heat rejection, or reactivity insertion. The consequences of single and double-fault accidents can be modeled, including fuel and coolant heating, fuel and cladding mechanical behavior, core reactivity feedbacks, coolant loop performance including natural circulation, and decay heat removal. Analyses are typically terminated upon demonstration of reactor and plant shutdown to permanently coolable conditions or upon violation of design basis margins. The objective of the analysis is to quantify accident consequences as measured by the transient behavior of system performance parameters, such as fuel and cladding temperatures, reactivity, and cladding strain. Originally developed for analysis of sodium cooled reactors with oxide fuel clad by stainless steel, the models In SAS4A/SASSYS-1 were subsequently extended and specialized to metallic fuel clad with advanced alloys and to several other coolant options including lead and LBE. METHOD OF SOLUTION: Each subassembly in the reactor core is represented with single or multiple-pin models. One channel represents one or more fuel pins and many channels are employed for a whole-core representation. Heat transfer in each pin is modeled with a two-dimensional (r/z) heat conduction equation. Single and two-phase coolant thermal-hydraulics are simulated with a unique, one-dimensional (axial) multiple-bubble liquid metal boiling model. The transient fuel and cladding mechanical behavior models are integrated with fission product production,release, and transport models. UNIQUE FEATURES: The physical models in SAS4AJSASSYS-1 are highly detailed 1umerical representations of reactor accident conditions based on extensive laboratory and test reactor results. The models are specialized to liquid metal (sodium) cooled fast reactors with oxide or metallic fuel clad with stainless steel.«

Fanning, Thomas↗

DOE OSTI2013

High Temperature Steam Oxidation Testing of Candidate Accident Tolerant Fuel Cladding Materials

The Fuel Cycle Research and Development (FCRD) program’s Advanced Fuels Campaign has initiated a multifold effort aimed at facilitating development of accident tolerant fuels in order to overcome the inherent shortcomings of light water reactor (LWR) fuels when exposed to beyond design basis accident conditions. The campaign has invested in development of experimental infrastructure within the Department of Energy complex capable of chronicling the performance of a wide range of concepts under prototypic accident conditions. This report summarizes progress made at Oak Ridge National Laboratory (ORNL) and Los Alamos National Laboratory (LANL) in FY13 toward these goals. Alternative fuel cladding materials to Zircaloy for accident tolerance and a significantly extended safety margin requires oxidation resistance to steam or steam-H 2 environments at ≥1200°C for short times. At ORNL, prior work focused attention on SiC, FeCr and FeCrAl as the most promising candidates for further development. Also, it was observed that elevated pressure and H2 additions had minor effects on alloy steam oxidation resistance, thus, 1 bar steam was adequate for screening potential candidates. Commercial Fe-20Cr-5Al alloys remain protective up to 1475°C in steam and CVD SiC up to 1700°C in steam. Alloy development has focused on Fe-Cr-Mn-Si-Y and Fe-Cr-Al-Y alloys with the aluminaforming alloys showing more promise. At 1200°C, ferritic binary Fe-Cr alloys required ≥25% Cr to be protective for this application. With minor alloy additions to Fe-Cr, more than 20%Cr was still required, which makes the alloy susceptible to α’ embrittlement. Based on current results, a Fe-15Cr-5Al-Y composition was selected for initial tube fabrication and welding for irradiation experiments in FY14. Evaluations of chemical vapor deposited (CVD) SiC were conducted up to 1700°C in steam. The reaction of H 2 O with the alumina reaction tube at 1700°C resulted in Al(OH) 3 formation, which reacted with the specimen to form a liquid reaction product. This behavior was not observed at lower temperatures where gas velocity and H 2 O content showed typical effects on the reaction rate. For LANL, the capabilities for oxidation testing as well as exploration of a methodology for measurement of hydrogen production of samples during oxidation under water vapor atmospheres is discussed. Results obtained for available commercial alloys are summarized, and data highlighting the performance of molybdenum, a recently proposed cladding material, are presented. Finally, leveraging of these techniques in conjunction with current and companion FCRD programs is discussed with respect to work in FY14.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2012

Assessment of Possible Cycle Lengths for Fully-Ceramic Micro-Encapsulated Fuel-Based Light Water Reactor Concepts

The tri-isotropic (TRISO) fuel developed for High Temperature reactors is known for its extraordinary fission product retention capabilities [1]. Recently, the possibility of extending the use of TRISO particle fuel to Light Water Reactor (LWR) technology, and perhaps other reactor concepts, has received significant attention [2]. The Deep Burn project [3] currently focuses on once-through burning of transuranic fissile and fissionable isotopes (TRU) in LWRs. The fuel form for this purpose is called Fully-Ceramic Micro-encapsulated (FCM) fuel, a concept that borrows the TRISO fuel particle design from high temperature reactor technology, but uses SiC as a matrix material rather than graphite. In addition, FCM fuel may also use a cladding made of a variety of possible material, again including SiC as an admissible choice. The FCM fuel used in the Deep Burn (DB) project showed promising results in terms of fission product retention at high burnup values and during high-temperature transients. In the case of DB applications, the fuel loading within a TRISO particle is constituted entirely of fissile or fissionable isotopes. Consequently, the fuel was shown to be capable of achieving reasonable burnup levels and cycle lengths, especially in the case of mixed cores (with coexisting DB and regular LWR UO2 fuels). In contrast, as shown below, the use of UO2-only FCM fuel in a LWR results in considerably shorter cycle length when compared to current-generation ordinary LWR designs. Indeed, the constraint of limited space availability for heavy metal loading within the TRISO particles of FCM fuel and the constraint of low (i.e., below 20 w/0) 235U enrichment combine to result in shorter cycle lengths compared to ordinary LWRs if typical LWR power densities are also assumed and if typical TRISO particle dimensions and UO2 kernels are specified. The primary focus of this summary is on using TRISO particles with up to 20 w/0 enriched uranium kernels loaded in Pressurized Water Reactor (PWR) assemblies. In addition to consideration of this 'naive' use of TRISO fuel in LWRs, several refined options are briefly examined and others are identified for further consideration including the use of advanced, high density fuel forms and larger kernel diameters and TRISO packing fractions. The combination of 800 {micro}m diameter kernels of 20% enriched UN and 50% TRISO packing fraction yielded reactivity sufficient to achieve comparable burnup to present-day PWR fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

The SAS4A/SASSYS-1 Version 5.8 Safety Analysis Code System

SAS4A/SASSYS-1 is a software simulation tool used to perform deterministic analysis of anticipated events as well as design basis and beyond design basis accidents for advanced nuclear reactors. Detailed, mechanistic models of steady-state and transient thermal, hydraulic, kinetic, and mechanical phenomena are employed to describe the response of the reactor core, the reactor primary and secondary coolant loops, the reactor control and protection systems, and the balance-of-plant to accidents caused by changes in coolant flow, loss of heat rejection, or reactivity insertion. The consequences of single and double-fault accidents can be modeled, including fuel and coolant heating, fuel and cladding mechanical behavior, core reactivity feedbacks, coolant loop performance including natural circulation, and decay heat removal. Analyses are typically terminated upon demonstration of reactor and plant shutdown to permanently coolable conditions, or upon violation of design basis margins. The objective of the analysis is to quantify accident consequences as measured by the transient behavior of system performance parameters, such as fuel and cladding temperatures, reactivity, and cladding strain. Originally developed for analysis of sodium cooled reactors with oxide fuel clad by stainless steel, the models were subsequently extended and specialized to metallic fuel clad with advanced alloys and to several other coolant options, including lead, LBE, and water.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗