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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 829 records · Page 46
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Film condensation with high heat fluxes and scaled experiments using pure steam for reactor containment cooling

Condensation tests were performed using a newly developed test facility for scaling the passive containment cooling system (PCCS) to a small modular reactor (SMR). The PCCS of the SMR plays a pivotal role in ensuring greater safety, reliability, and compactness than what is afforded by traditional reactors. Therefore, a well-designed PCCS is essential to SMRs. However, previous studies and test data were unsuitable for scaling, due to high variation in the test geometry and operating conditions. This study intends to close this research gap by using a novel designed scaled test facility consisting of vertical condensing test sections featuring 1-, 2-, and 4-inch-diameter condensing tubes with annular water cooling, and by applying superheated and saturated steam with different steam mass flow ranges of 5–25 g/s. Further, the primary test data, including axial temperatures, mass flow rates, and pressures, were used in conjunction with a standard data reduction method to estimate critical parameters such as heat fluxes, heat transfer coefficients, and condensation rates. These scaled test data would support improving empirical correlations and validating condensation models to identify scaling distortion for SMR PCCSs.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Status of the OECD/NEA Watts Bar unit 1 benchmark and calculations of local reactor core power data of the Zero Power physics tests

The paper aims to provide an update on the status of the Organization for Economic Cooperation and Development (OECD) / the Nuclear Energy Agency (NEA) Tennessee Valley Authority (TVA) Watts Bar 1 (WB1) Multi-Physics Multi-Cycle depletion benchmark and the results associated with Exercise 1. The benchmark relies on the set of benchmark progression problems developed by the Department of Energy (DOE) Consortium for the Advanced Simulation of Light Water Reactors (CASL) for the Virtual Environment for Reactor (VERA), which are based on real plant design and operational data. The OECD/NEA TVA WB1 benchmark is designed for validation of both traditional and novel high-fidelity multi-physics codes to analyze Pressurized Water Reactors (PWR) depletion cycles. The activities are conducted under the Expert Group on Reactor Systems Multi-Physics (EGMUP) at NEA/OECD. In this work, we analyze Exercise 1 of the benchmark: stand-alone Three-Dimensional (3D) neutronics at Start-up Zero Power Physics Test (ZPPT) at Hot Zero Power Conditions (HZP) and the resulting power maps. The code used to model the exercise is the continuous energy Monte Carlo code Serpent 2.1.31 along with the nuclear data library ENDF/B-VII.1. Serpent results are compared with the high-fidelity deterministic code MPACT, which is part of VERA. The paper presents three selected results from power calculations required output. The results compared are the normalized axially integrated radial core power maps, the normalized axial averaged core power shapes, and the normalized core hottest and coldest assemblies radial power maps. The Root Mean Square Deviation (RMSD) between Serpent and MPACT is 0.51 % for the axially integrated radial core power map, 1.41 % for the axial averaged core power shape, 2.42 % for the hottest assembly axial power shape, 0.88 % for the coldest assembly axial integrated hottest assembly radial pin power map, and 0.17 % for the axially integrated coldest assembly radial pin power map. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

SPC-2918 Rev 0 MARVEL Test Loop Flowmeters

This specification contains the requirements for the design, manufacture, testing, and delivery of magnetic flowmeters to be installed on the MARVEL Primary Coolant Apparatus Test (PCAT) system. The PCAT system is a heated test loop which will be used to simulate the coolant system of a proposed sodium or NaK cooled microreactor. The system has four identical cooling loops. Installation of flowmeters on all four loops is anticipated at this time, however, the number of flowmeters may be reduced by the Project. Note that the number of flowmeters described herein is not a commitment to a particular number of procured units. The number of flowmeter units procured shall be identified on the purchase order. The MARVEL reactor is still under design, so the PCAT flowmeter design may not be directly transferrable to the reactor. It is intended that the design and development effort for the PCAT flowmeters be leveraged to the extent practical for the reactor flowmeter design.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2020

Scoping Analysis of Sodium Cooled Fast Spectrum Test Reactor Cores

A team from the national laboratories conducted a study that assessed advanced reactor technology options and the irradiation needs of a broad group of stakeholders. This identified a gap in the fast neutron irradiation capability - the need for a test reactor that can deliver high neutron fluxes of fast neutrons. The study compared fast and thermal neutron irradiation technology options based on technology maturity and the ability to meet strategic objectives, along with pathways to deployment and tradeoffs in mission. It also provided a recommendation on specific technology options to address the current gap in fast neutron irradiation capabilities. In this context, a detailed scope of work was prepared that pertains to a trade study focusing on the core of a fast neutron spectrum test reactor. This report presents the main results obtained by a team of Argonne National Laboratory (ANL) and Idaho National Laboratory (INL) scientists. It aims at developing an overall understanding of how the various areas important to the design and successful operation of such a facility interact with one another. Developing a sense of the relevant sensitivities at a systems level creates an understanding of where and how margins may be available to improve performances (e.g., flux level, capacity factor, and thermal hydraulic limits). Emphasis is given to the reactor (core design options and safety) and its fuel cycle (fuel supply and fabrication, as well as spent fuel management).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Uncertainty and Sensitivity Analysis of TREAT Transient Test no. 15

The U.S. Department of Energy (DOE) is preparing for the resumption of transient testing using the Transient Reactor Test Facility (TREAT) located at the Idaho National Laboratory (INL). In order to support the analysis of TREAT operation and experiment planning, the MAMMOTH reactor physics team has recently begun development of uncertainty and sensitivity analysis for the TREAT core simulated via MAMMOTH. MAMMOTH is a high-fidelity three-dimensional transient multi-physics code developed at INL under the DOE-NE Advanced Modeling and Simulation program (NEAMS). This research is part of the development presenting some initial uncertainty and sensitivity analysis of TREAT Transient Test no. 15. In this summary, we employ the Sobol sampling strategy implemented in RAVEN to efficiently perform the sensitivity and uncertainty analysis. The idea is using Sobol functional decomposition with generalized polynomial chaos (gPC) expansions to estimate the responses of interest. The gPC coefficients are then computed via the sparse grid spectral projection method. In addition, this strategy includes evaluation of the second or higher order sensitivity indices, allowing the study of interactions between several input variables. The principal focus of this analysis is to quantify the uncertainties in the total power and maximum fuel temperatures with respect to uncertainties in the initial transient conditions, such as rated power, fuel temperature, graphite fuel density and control rod boron concentration. In addition, both the first order and second order time-dependent sensitivity indices are computed and analyzed. The results show that the interactions between initial control rod boron concentration and the rated power will contribute at most 1% to the total uncertainty of the power peak.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Scaling Analysis of Vortex Shedding Frequency between In line Plates

The Global Threat Reduction initiative (GTRI) was established by the National Nuclear Security Administration (NNSA) in 2004. One of the highlights of the program is to convert research and test reactors at home and abroad from Highly Enriched Uranium (HEU) to Low Enriched Uranium (LEU) fuels. By switching directly from HEU to LEU, an LEU reactor produces fewer neutrons than an HEU reactor and in some cases; a direct switch from HEU to LEU results in the reactor being unable to achieve criticality. Under certain circumstances, criticality is achieved and the original neutron fluxes are restored with LEU fuel by adding more fuel elements to the core. Not all reactors, exemplified by the Advanced Test Reactor (ATR) at the Idaho National Lab (INL), have the luxury of simply adding more fuel to the core. In such cases, entirely new designs must be created and tested. The ATR utilizes plate type fuel. During the design and testing of the ATR fuel, vortex shedding was identified as potential source for experiment and fuel failure. For fuel conversion, elements must be tested in pile prior to the qualification of the new fuel design. For some test elements, the primary test plate is followed by a secondary handling plate. It is known that objects downstream influence the vortex shedding. While vortex shedding is an important and well-studied phenomenon, not only in the context of nuclear engineering, but in fields as diverse as aeronautical engineering and biology, there is a significant gap in the understanding of vortex shedding in regards to two bluff bodies in line direct line with one another. The study provides a scaling analysis of the frequency of vortex shedding for plates in line. The scaling analysis is then compared to numerical simulation of two in line plates. The study provided a scaling analysis and suggested parameter for determining the expected trends in the Strouhal number for a plate with some object downstream. The analysis made the consideration of differing length scales as the cause for a reduction in vortex shedding frequency as well as looked at the potential impact of boundary layer growth on vortex shedding for a flat plate. The comparison between the expected scaling and numerical data was less than ideal, but there are several potential causes. One of the biggest issues with the comparison is the uncertainty associated with the defect velocity. The frequency was well established in several cases, but the velocity was not always well established resulting in a larger uncertainty. In the cases where the velocity was near zero (i.e. b/d = 0.5), the uncertainty dominated since the corrected Strouhal number is divided by the defect velocity. The Reynolds numbers were not incredibly large, ranging from 350-1600. The low Reynolds numbers potentially undermine the principle assumption made in the scaling analysis, i.e. Reynolds number is negligibly small. Additionally, the exact definition of the 'reduced velocity' was never explicitly defined, merely inferred from the assumption that the x length scale and y length scale were in fact different. Another potential problem is the defect velocities in the lower Reynolds number cases were not much smaller than the free stream velocity. Further work needs to be done at higher Reynolds numbers and with more data points to ensure the assumptions made are valid as well as confirm the accuracy of the scaled relationship. (authors)

42 ENGINEERING↗
DOE OSTI2016

Investigation of similarity metrics for simulation based scaling analysis

Scaling analysis is a technique for characterizing degree of similarity between processes to occur in a real prototypic system and those to occur in its reduced scale (experimental) analog. In the context of nuclear reactor thermal-hydraulics and safety, scaling analysis is instrumental in assessing the applicability of data obtained from a scaled test facility for calibration of models and validation of simulation codes. Quantification of scale-related distortions is the key element in scaling analysis. In this work, we examine an approach for scaling analysis using computer simulation codes. Similarity analysis using information-theoretic approach is performed by comparing performance of experiment and application to variation of system conditions. Three different information theoretic measures namely mutual information, Kullback-Leibler divergence and overlap coefficient are selected for their potential use as metric for similarity analysis. We test the metrics by comparing them against physics-based scaling ratios, which are well known in fundamental thermal-hydraulic processes. In this exploratory analysis, a case of pressure drop of a steady-state flow through a channel is employed to test the proposed approach. It was found that in the absence of model uncertainties, mutual information serve as an effective measure of the scale effects present in the experiment. However, model uncertainty can significantly impact the quality of similarity analysis. Our analysis indicates that model uncertainties can have significant impact on the information theoretic metrics used for scaling analysis. Furthermore, if the model uncertainties exceeds beyond a critical value, the scale effect present in the experiment cannot be adequately evaluated. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Down-selection of candidate alloys for further testing of advanced replacement materials for LWR core internals

Life extension of the existing nuclear reactors imposes irradiation of high fluences to structural materials, resulting in significant challenges to the traditional reactor materials such as type 304 and 316 stainless steels. Advanced alloys with superior radiation resistance will increase safety margins, design flexibility, and economics for not only the life extension of the existing fleet but also new builds with advanced reactor designs. The Electric Power Research Institute (EPRI) teamed up with Department of Energy (DOE) Light Water Reactor Sustainability Program to initiate the Advanced Radiation Resistant Materials (ARRM) program, aiming to identify and develop advanced alloys with superior degradation resistance in light water reactor (LWR)-relevant environments by 2024.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Analysis on the Preliminary Gap Conductance Test with Single Primary Heat Extraction and Removal Emulator

Microreactor is a type of nuclear reactor that is intended to produce thermal energy from the micro and integrated design. Among various microreactor roadmaps, microreactor with heat pipes is one of the promising designs that under active development. At Idaho National Laboratory (INL) and supported by the Department of Energy (DOE) Microreactor Program (MRP), a Single Primary Heat Extraction and Removal Emulator (SPHERE) was developed. The main purpose of SPHRE is to understand thermal performance of heat pipe in the conjunction of heat source without using nuclear material. The heat pipe used in the SPHERE has sodium as a working fluid, with Argon gas in the vapor region, thus this heat pipe is variable conductance heat pipe (VCHP). The volume of non-condensable gas changes with power and elapsed time, so the area covered by non-condensable gas changes. This gives an effect that passively controls the cooling capability of condenser section. This preliminary experiment was intended to perform a gap conductance experiment that reveals heat transfer characteristics of gas at the gap around the heat pipe. However, there are not many experimental data that deals with heat transfer of the gas at the gap around heat pipe. In this research, the preliminary experimental result regarding to the heat pipe performance with different composition of gas at the gap, level of power is presented. The experimental data will be used for the validation study of SOCKEYE as a future plan.

42 ENGINEERING↗
DOE OSTI2017

Development and Test of a New Verification Scheme for Transient Core Simulators

Transient calculations in commercial nuclear reactors are performed while typically relying on a time-dependent neutron transport solver or a low-order solver (i.e. diffusion). In order to be licensed, the codes used by the industry need to go through a process of verification and validation, with the verification carried out by comparing the results of simulations to analytical or semi-analytical solutions. Such analytical or semi-analytical solutions can only be obtained if the system to be modelled during the verification process is either fully homogeneous or piecewise homogeneous. This paper reports on the development of a different verification approach that can be applied to fully heterogeneous systems. It relies on the extraction of the point-kinetic response of the reactor (which can be estimated from the results of core simulations) and on its subsequent comparison with its expected analytical form.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Initial Results from the High Fluence UCSB ATR-2 Irradiation

A primary focus of the Department of Energy, Office of Nuclear Energy Light Water Reactor Sustainability Program (LWRSP) is the safe operation of nuclear reactors to extended lifetimes of up to 80 years. One of the primary reactor components of concern is the reactor pressure vessel (RPV), which is increasingly embrittled over time due to damage from high energy neutrons created in the reactor core. Radiation damage leads to the formation of nm-scale precipitates and solute-defect clusters, which result in an increase in the steel's ductile to brittle transition temperature. Transition temperature shifts (TTS) are used to quantify and regulate embrittlement. Extended life operation will subject some RPVs to higher neutron fluences than previously experienced by in-service vessels. Thus, robust prediction methods must be developed that ensure the safety of our nation's fleet of reactors. Current models underpredict embrittlement data from high flux test reactor irradiations at extended life fluences. It has been hypothesized that one reason for this underprediction is the formation of so-called 'Late- Blooming Phase' Mn-Ni-Si precipitates, which form and grow only at high fluence and that are currently not included in regulatory models. Recent research has demonstrated the existence of high volume fractions of intermetallic Mn-Ni-Si precipitate (MNSP) phases at high fluence and have generated significant insight into their character. However, a number of questions remain regarding MNSPs, such as: 1) what alloys are most susceptible to MNSP formation; 2) how is MNSP formation affected by flux (φ), fluence (φt) and irradiation temperature (T{sub irr}); and, 3) how to accurately predict TTS under extended life low flux service conditions? The fact that no two RPVs are exactly the same, both in terms of the steel composition and vessel service conditions, makes these questions particularly difficult to answer. Further, understanding the effect of flux differences between test reactor irradiations and in-vessel service is absolutely critical. Hence, physical models must be developed that predict TTS as a function of alloy solute content and vessel service conditions. A large-scale irradiation designed to develop and validate an extended life embrittlement prediction model was recently completed in the Advanced Test Reactor (ATR) at the Idaho National Laboratory as part of the DOE sponsored Nuclear Scientific User Facilities. This experiment, called the UCSB ATR-2 irradiation, is a collaboration between Oak Ridge National Lab (ORNL) and the University of California, Santa Barbara (UCSB) and is now primarily supported by the DOE Light Water Reactor Sustainability Program. The experiment is designed to investigate the single and combined effects of neutron φ, φt, T{sub irr} and alloy chemistry on precipitate formation, hardening and embrittlement. While postirradiation examinations (PIE) is ongoing, and is expected to take several years to complete, the first preliminary results are presented below. It should also be noted that Rolls-Royce has been a collaborator in UCSB ATR-2 from the start, and there are several other national and international institutions supporting the PIE. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2021

MITR Low-Enriched Uranium Conversion Fluid-Structure Interaction Preliminary Design Verification

The U.S. National Nuclear Security Administration (NNSA) Office of Material Management and Minimization (M 3 ) has developed and is pursuing an integrated approach to address the persistent threat posed by unintentional proliferation of nuclear materials. The NNSA M 3 approach reduces the risk of highly enriched uranium (HEU) and plutonium falling into the hands of non-state actors by minimizing the use of and, when possible, eliminating weapons-usable nuclear material around the world. In this geopolitical context, most research and test reactors, both domestic and international, have completed or started a program of conversion from the use of HEU to low-enriched uranium (LEU) as fuel. Conversion of civilian research reactors from HEU to LEU, and the return of the HEU to the country of origin, are important components of the NNSA non-proliferation program. Worldwide, 71 reactors have been converted to the use of LEU fuel, and an additional 32 have been confirmed to be permanently shut down. The U.S., with 20 reactor conversions, is among the 39 countries on six continents where conversions have occurred. With recent conversions in Ghana and Nigeria, an important milestone was reached in completing conversion of all reactors on the continent of Africa to LEU fuel. Africa thus becomes the third continent to have completed conversion of all HEU reactors to LEU, following Australia and South America

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Benchmark Study of Infinite Lattice of Pebble Fuel within FLiBe Coolant for TMSR-SF1 Design

Solid Fueled Thorium Molten Salt Reactor (TMSR-SF1) is Chinese Fluoride-salt-cooled High Temperature Reactor (FHR) designed by the TMSR group. The construction of this test reactor is expected in the near years. Using TRISO particle based pebble fuel leads to various challenges in neutronics modeling, such as double heterogeneity effect, preserving exact number of particles, their random distributions in pebble, and etc. For the purpose of verifying the calculations, benchmark study of infinite lattice of pebble fuel within FLiBe coolant for the TMSR-SF1 design has been performed. Back-to-back comparisons have been made between TMSR and MIT researchers using different Monte-Carlo based codes. Very good agreements have been found and the results quantitatively evaluate the neutronic effect of various modeling approaches. In addition, sensitivity study has been carried out for the reference pebble-bed fuel lattice. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Nuclear Thermal Rocket Emulator for a Hardware-in-the-Loop Test Bed

To support NASA’s mission to use nuclear thermal rockets for future Mars missions, an instrumentation and control test bed has been built at Oak Ridge National Laboratory. The system is designed as a hardware-in-the-loop test bed for testing control elements and autonomous control algorithms for nuclear thermal propulsion rockets. The mock reactor system consists of a modular and scalable framework, using inexpensive components and open-source software. The hardware system consists of a two-phase flow loop and a mock reactor with six control drums. A single-board computer (NVIDIA Jetson) handles reactor core emulation and hosts a message queuing telemetry transport broker that allows user-deployed control algorithms to interact with the system hardware. The reactor emulator receives sensor data from the hardware and provides the simulated performance of the reactor under steady-state, transient, and fault conditions. The emulator uses a reactivity lookup table and the point kinetics equations to solve for the reactor dynamics in real time. Emulated reactor dynamics and sensor input inform the autonomous control algorithm’s decision-making in a closed-loop manner. The current system is capable of operating at 10 Hz, but faster cycle rates are an area of ongoing research. This test bed will enable NASA and other space vendors to rigorously test their autonomous control systems for NTP rockets under transient (reactor startup and shutdown), steady-state, and fault conditions to reduce development time and risk for autonomous control systems in future missions.

autonomous control↗
DOE OSTI2023

Non-Nuclear Test Facility Development to Support Verification and Validation Needs

This study focuses on developing a non-nuclear test facility to help meet the verification and validation (V&V) requirements of a water-cooled small modular reactor (SMR) system. The scaling analysis necessary for reactor system experimentation and assessment model V&V considers various reactor system accident scenarios such as main steam line break (MSLB), steam generator tube rupture (SGTR), and containment condensation and heat transfer. The goal is to identify key physics based phenomena of interest (POI), specific figures of merit (FOMs), and the state of knowledge (SOK) by employing phenomena identification and ranking table (PIRT) studies. Important physics phenomena such as heat transfer, fluid dynamics, structural integrity, and material selection must be addressed when scaling a thermal-hydraulics system. Alongside these considerations, modifications to the facility may be needed to align the scaled system with the available footprint (height and space) of the facility. These modifications may include infrastructure upgrades, changes to the piping layout, implementation of safety measures, and installation of new instrumentation and control systems. The test facility aims to provide a scaled-down version of prototypic facilities for comprehensive V&V activities by overcoming these challenges and making the necessary modifications. This approach minimizes scaling distortion, simulates prototypic plant conditions, and accurately models accident event progressions. Thus, the facility will be capable of adequately modeling a scaled SMR system to support the design, development and licensing needs.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Overlapping Domain Coupling of Multidimensional and System Codes in NEAMS - Pronghorn and SAM

This report describes the implementation and testing of domain-overlapping coupling of the Pronghorn and SAM codes. Both Pronghorn and SAM are codes developed by the DOE Nuclear Energy Advanced Modeling and Simulation (NEAMS) program for analysis of advanced nuclear reactors. Pronghorn focuses on analyzing multi-dimensional core flow conditions, while SAM focuses on analysis of the entire system including, among other components, piping, pumps, and heat exchangers. It is desirable for many advanced reactor thermal-hydraulics simulations to couple Pronghorn and SAM to obtain self-consistent solutions in the core and the system. The domain-overlapping coupling approach provides a robust and numerically efficient candidate for this coupling. We describe the algorithm for obtaining self-consistent solutions between a SAM network of flow channels and an overlapped multi-dimensional Pronghorn domain. The algorithm is tested for a variety of simple canonical test problems and a model of the Molten Salt Reactor Experiment (MSRE). Results demonstrate that the domain overlapping coupling approach provides consistent pressure, mass flow rate, enthalpy, and passive scalar distributions between Pronghorn and SAM and converges reliably and efficiently (at most 15 iterations, usually much less than 10) for the considered test problems.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2016

TREAT neutronics analysis and design support, Part II: Multi-SERTTA-CAL

Experiment vehicle design is necessary in preparation for Transient Reactor Test (TREAT) facility restart and the resumption of transient testing to support Accident Tolerant Fuel (ATF) characterization and other future fuels testing requirements. Currently the most mature vehicle design is the Multi-SERTTA (Static Environments Rodlet Transient Test Apparatuses), which can accommodate up to four concurrent rodlet-sized specimens under separate environmental conditions. Robust test vehicle design requires neutronics analyses to support design development, optimization of the power coupling factor (PCF) to efficiently maximize energy generation in the test fuel rodlets, and experiment safety analyses. An integral aspect of prior TREAT transient testing was the incorporation of calibration experiments to experimentally evaluate and validate test conditions in preparation of the actual fuel testing. The calibration experiment package established the test parameter conditions to support fine-tuning of the computational models to deliver the required energy deposition to the fuel samples. The calibration vehicle was designed to be as near neutronically equivalent to the experiment vehicle as possible to minimize errors between the calibration and final tests. The Multi-SERTTA-CAL vehicle was designed to serve as the calibration vehicle supporting Multi-SERTTA experimentation. Models of the Multi-SERTTA-CAL vehicle containing typical PWR-fuel rodlets were prepared and neutronics calculations were performed using MCNP6.1 with ENDF/B-VII.1 nuclear data libraries; these results were then compared against those performed for Multi-SERTTA to determine the similarity and possible design modification necessary prior to construction of these experiment vehicles. The estimated reactivity insertion worth into the TREAT core is very similar between the two vehicle designs, with the primary physical difference being a hollow Inconel tube running down the length of the calibration vehicle. Calculations of PCF indicate that on average there is a reduction of approximately 6 % for PWR fuel rodlets irradiated under both wet and dry conditions. Changes to the primary or secondary vessel structure in the calibration vehicle can be performed to offset this discrepancy and maintain neutronic equivalency. Current possible modifications to the calibration vehicle include reduction of the primary vessel wall thickness, swapping Zircaloy-4 for stainless steel 316 in the secondary containment, or slight modification to the temperature and pressure of the water environment within the primary vessel. Removal of some of the instrumentation within the calibration vehicle can also serve to slightly increase the PCF. Future efforts include further modification and optimization of the Multi-SERTTA and Multi-SERTTA-CAL designs in preparation of actual TREAT transient testing. Experimental results from both test vehicles will be compared against calculational results and methods to provide validation and support additional neutronics analyses. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗