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DOE OSTI2014

Early User Test: BISON for Fission Gas Release Modeling

As a test to develop an understanding of the capabilities and input of the nuclear fuel performance code BISON, we modeled the fission gas release of the Oconee rod 15309 as there is benchmark data available for comparison. BISON has two fission gas models available, the Forsber-Massih model and the Simple Integrated Fission Gas Release and Swelling [sifgrs] model. Due to time constraints and availability of model parameters [we did not have all of the parameters for the sifgrs model already in hand] we only tested the Forsberg-Massih model. The Oconee rod test has been used in a number of benchmark studies including the integral assessment of FRAPCON 3.4. The Oconee Rod test is a 5-cycle test bundle with an average burnup of 50GWd/MTU. The FRAPCON assessment predicted that Oconee Rod15309 Test case would release 1.25% of all fission gas produced, while experimentally only 0.8% was released at EOL. Other studies using BISON with the sifgrs model have under predicted EOL fission gas release [FGR].

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Non-Destructive Evaluation of Polyolefin Thermal Aging Using Infrared Spectroscopy

Fourier transform infrared (FTIR) spectroscopy is an information-rich method that reveals chemical bonding near the surface of polymer composites. FTIR can be used to verify composite composition, identify chemical contaminants and expose composite moisture content. Polymer matrix changes due to thermal exposure including loss of additives, chain scission, oxidation and changes in crystallinity may also be determined using FTIR spectra. Portable handheld instruments using non-contact reflectance or surface contact attenuated total reflectance (ATR) may be used for non-destructive evaluation (NDE) of thermal aging in polymer and composite materials of in-service components. We report the use of ATR FTIR to track oxidative thermal aging in ethylene-propylene rubber (EPR) and chlorinated polyethylene (CPE) materials used in medium voltage nuclear power plant electrical cable insulation and jacketing. Mechanical property changes of the EPR and CPE materials with thermal degradation for correlation with FTIR data are tracked using indenter modulus (IM) testing. IM is often used as a local NDE metric of cable jacket health. The FTIR-determined carbonyl index was found to increase with IM and may be a valuable NDE metric with advantages over IM for assessing cable remaining useful life.

aging↗
DOE OSTI2024

Sister Rod Destructive Examinations (FY23) Appendix I: SNF Aerosols Released During Rod Fracture

This report documents work performed under the Spent Fuel and Waste Disposition’s Spent Fuel and Waste Science and Technology program for the US Department of Energy (DOE) Office of Nuclear Energy (NE). This work was performed to fulfill Level 3 Milestone M3SF-23OR010201026, “FY23 M3 draft report on results from testing in FY23,” within work package SF-23OR01020102 and is an updated to the work reported in M2SF-23OR010201024, M2SF-22OR010201047, M2SF-21OR010201032, M2SF-19ORO010201026, and M2SF-19OR010201028. As a part of the DOE NE High Burnup Spent Fuel Data Project, Oak Ridge National Laboratory (ORNL) is performing destructive examinations (DEs) of high burnup (HBU) (>45 GWd/MTU) spent nuclear fuel (SNF) rods from the North Anna Nuclear Power Station operated by Dominion Energy. The SNF rods, called sister rods or sibling rods, are all HBU and include four different kinds of fuel rod cladding: standard Zircaloy-4 (Zirc-4), low-tin Zirc-4, ZIRLO, and M5. The DEs are being conducted to obtain a baseline of the HBU rods’ condition before dry storage and are focused on understanding overall SNF rod strength and durability. Fuel rods and defueled cladding will be tested to derive material properties. Although the data generated can be used for multiple purposes, one primary goal for obtaining the post-irradiation examination data and the associated measured mechanical properties is to support SNF dry storage licensing and relicensing activities by (1) addressing identified knowledge gaps and (2) enhancing the technical basis for post-storage transportation, handling, and subsequent disposition.This report documents the status of the ORNL Phase 1 DE activities related to the collection of SNF aerosol particles released during fuel rod fracture in 4-point bending in Phase 1 of the sister rod test program.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

AGR Irradiated Specimen Air/Moisture Heating Test Plan

The Advanced Gas Reactor (AGR) program has a need to perform high-temperature oxidation testing of tristructural isotropic (TRISO)-coated particle fuel and reactor core materials to evaluate their performance under conditions relevant to air- and moisture-ingress events in high-temperature gas-cooled reactors (HTGRs). The objectives of these tests are to assess the impact of oxidizing environments on fission product release from exposed fuel kernels, compact matrix, and nuclear grade graphite, and to determine the impact of oxidizing environments on TRISO coating integrity. The program is currently developing a capability for testing irradiated fuel, matrix, and graphite specimens inside a hot cell. The Air/Moisture Ingress Experiment (AMIX) system will be capable of temperatures up to 1650°C in atmospheres containing oxygen or moisture with varying concentrations. The system will be capable of collecting fission gases as well as condensable fission products, in order to quantify total released inventory from the heated specimens. A basis for establishing experimental test matrices, and the test matrices themselves, are presented in this document. This includes considerations on specific objectives for the tests, types of oxidizing environments that should be tested, and the expected availability of irradiated specimens from the various AGR program irradiation experiments. This document only briefly discusses post-test analysis of the heated specimens, and it is recognized that in many cases the post-test analysis may be extensive. It is expected that this will be a “living” document, subject to revision as data are collected and scoping calculations are performed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

US-ROK Action Sheet 34: Safeguards Application of a Hand-held Mechanically Cooled Germanium Spectrometer

This report summarizes results of Action Sheet 34 - for the cooperative efforts on the field testing and evaluation of a high-resolution, hand-held, gamma-ray spectrometer, known as SPG (Spectroscopic Planar Germanium), for safeguards application such as short notice inspections, UF6 analysis, enrichment determination, and other potential applications. The Spectroscopic Planar Germanium (SPG) has been demonstrated IAEA Physical Inventory Verification (PIV) in South Korea. This field test was a success and the feedback provided by KINAC, IAEA, and national laboratory staff was used to direct efforts to improve the instrument this year. Key points in this report include measurement results from PIV, analysis of spectra with commercially available Ortec U235 and PC-FRAM, and completion of tripod and tungsten collimator and integration of user feedback.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2017

Frequency spectrum analysis code for nuclear noise based on improved wavelet algorithm - 337

The core of the narrow flow channel generally faced the problem of local bubble boiling with the temperature rise of the nuclear fuel element, it is necessary and meaningful to answer and deal with the problem on the technical level. Reactor noise analysis begins with the study of the physical characteristics of the core, after nearly 30 years of research and development it has become a powerful tool in the auxiliary monitoring of the power reactor operating. By using this noise frequency spectrum analysis technology, the reactor dynamic parameter such as prompt neutron decay constant can be measured. Also, through such system, the frequency characteristics of reactor can be real-time monitored, which is helpful for safety operation and fault diagnosis. Spectrum analysis technology in the Fourier analysis era is now barely available, the reason is that the local spectral characteristics of this technology are not obvious, but this defect can be greatly improved by wavelet analysis technique, however, the traditional wavelet analysis will introduce the problem of spectral confusion. To solve the problem, based on the improved wavelet algorithm, frequency spectrum analysis code for nuclear noise is developed,from the perspective of fine spectral analysis, the local boiling accident of the coolant can be prejudged and avoid. Based on the analysis and test of experimental data,The results show that the calculation ability of the code is in accordance with the expected, it can be applied in the operation of research reactor. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Predictive based monitoring of nuclear plant component degradation using support vector regression

Nuclear power plants (NPPs) are large installations comprised of many active and passive assets. Degradation monitoring of all these assets is expensive (labor cost) and highly demanding task. In this paper a framework based on Support Vector Regression (SVR) for online surveillance of critical parameter degradation of NPP components is proposed. In this case, on time replacement or maintenance of components will prevent potential plant malfunctions, and reduce the overall operational cost. In the current work, we apply SVR equipped with a Gaussian kernel function to monitor components. Monitoring includes the one-step-ahead prediction of the component’s respective operational quantity using the SVR model, while the SVR model is trained using a set of previous recorded degradation histories of similar components. Predictive capability of the model is evaluated upon arrival of a sensor measurement, which is compared to the component failure threshold. A maintenance decision is based on a fuzzy inference system that utilizes three parameters: (i) prediction evaluation in the previous steps, (ii) predicted value of the current step, (iii) and difference of current predicted value with components failure thresholds. The proposed framework will be tested on turbine blade degradation data.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Bayesian Exploration of Phenomenological EoS of Neutron/Hybrid Stars with Recent Observations

The description of the stellar interior of compact stars remains as a big challenge for the nuclear astrophysics community. The consolidated knowledge is restricted to density regions around the saturation of hadronic matter ρ 0 =2.8 × 10 14 g cm -3 , regimes where our nuclear models are successfully applied. As one moves towards higher densities and extreme conditions up to the quark/gluons deconfinement, little can be said about the microphysics of the equation of state (EoS). Here, we employ a Markov Chain Monte Carlo (MCMC) strategy to access the variability at high density regions of polytropic piecewise models for neutron star (NS) EoS or possible hybrid stars, i.e., a NS with a small quark-matter core. With a fixed description of the hadronic matter for low density, below the nuclear saturation density, we explore a variety of models for the high density regimes leading to stellar masses near to 2.5 M ⊙ , in accordance with the observations of massive pulsars. The models are constrained, including the observation of the merger of neutrons stars from VIRGO-LIGO and with the pulsar observed by NICER. In addition, we also discuss the possibility of the use of a Bayesian power regression model with heteroscedastic error. The set of EoS from the Laser Interferometer Gravitational-Wave Observatory (LIGO) was used as input and treated as the data set for the testing case.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Fire-Induced Pressure Response and Failure Characterization of PCV/SCV/3013 Containers - Phase 3

Several Department of Energy (DOE) facilities have materials stored in robust, welded, stainless - steel containers with presumed fire - induced pressure response behaviors. Lack of test data related to fire exposure requires conservative safety analysis assumptions for container response at these facilities. This conservatism can in turn result in the implementation of challenging operational restrictions with costly nuclear safety controls. To help address this issue for sites that store DOE 3013 stainless - steel containers, a series of ten tests were undertaken at Sandia National Laboratories. The goal of this test series was to obtain the response behavior for various configurations of DOE 3013 containers with various payload compositions when exposed to one of two ASTM fire conditions. Key parameters measured in the test series included identification of failure - specific characteristics such as pressure, temperature, and whether or not a vessel was breached during a test . Numerous failure - specific characteristics were identified from the ten tests. This report describes the implementation and execution of the test series performed to identify these failure - specific characteristics. Discussions on the test configurations, payload compositions, thermal insults, and experimental setups are presented. Test results in terms of pressurization and vessel breach (or no - breach) are presented along with corresponding discussions for each test.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Modeling Astrophysical Explosions and the Nucleosynthesis of the Heavy Elements

The heavier elements in the Universe were created and dispersed in the supernova explosions of stars. These events involve a fascinating interplay of nuclear physics, gravity, hydrodynamics, neutrino and photon transport, and radioactive decay, all occurring under extreme conditions. Astronomical experiments provide a rich and constraining data set, but our theoretical understanding of supernovae remains incomplete. The goal of this project is to advance numerical models of astrophysical explosions and use them to connect the nuclear theory of supernovae to current observational and experimental programs. Our contributions focus on one important aspect of the problem --modeling the transport of radiation (both neutrino and electromagnetic) which is essential to understanding the supernova explosion mechanism, the resulting nucleosynthesis, and the observable spectra and light curves. We will exploit recent advances in high performance computing to improve 3-dimensional simulations of core collapse supernovae and neutron star mergers. The benefits include: 1) Further defining the explosive astrophysical environments that govern the nucleosynthesis of r-process nuclei. 2) Preparing the way for more realistic treatments of neutrino interactions, with the ultimate goal of making supernovae a testing ground for probing fundamental neutrino physics, e.g., unknown mixing angles and new aspects of the MSW mechanism. 3) Providing a means for testing explosion simulations and nucleosynthesis calculations against data from optical and x-ray telescopes, thereby permitting the experimental validation or falsification of supernova models. 4) Helping position nuclear physics, through the development of new algorithms of the type incorporated in our simulation codes, to play a leading role in DOE large scale computing.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

SNM Radiation Signature Classification Using Different Semi-Supervised Machine Learning Models

The timely detection of special nuclear material (SNM) transfers between nuclear facilities is an important monitoring objective in nuclear nonproliferation. Persistent monitoring enabled by successful detection and characterization of radiological material movements could greatly enhance the nuclear nonproliferation mission in a range of applications. Supervised machine learning can be used to signal detections when material is present if a model is trained on sufficient volumes of labeled measurements. However, the nuclear monitoring data needed to train robust machine learning models can be costly to label since radiation spectra may require strict scrutiny for characterization. Therefore, this work investigates the application of semi-supervised learning to utilize both labeled and unlabeled data. As a demonstration experiment, radiation measurements from sodium iodide (NaI) detectors are provided by the Multi-Informatics for Nuclear Operating Scenarios (MINOS) venture at Oak Ridge National Laboratory (ORNL) as sample data. Anomalous measurements are identified using a method of statistical hypothesis testing. After background estimation, an energy-dependent spectroscopic analysis is used to characterize an anomaly based on its radiation signatures. In the absence of ground-truth information, a labeling heuristic provides data necessary for training and testing machine learning models. Supervised logistic regression serves as a baseline to compare three semi-supervised machine learning models: co-training, label propagation, and a convolutional neural network (CNN). In each case, the semi-supervised models outperform logistic regression, suggesting that unlabeled data can be valuable when training and demonstrating value in semi-supervised nonproliferation implementations.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2026

Advanced Multiphysics Code Coupling for Cladding Surface Thermocouples During Two-Phase Heat Transfer from Nuclear Fuel

Transient testing of nuclear fuel involves the evaluation of fuel performance under off-normal and accident conditions and is essential for proving design performance. Instrumentation included in such experiments commonly includes thermocouples attached to the outer surface of the cladding to provide temperature measurements throughout the transient. However, the presence of thermocouples on the cladding surface can alter the local heat transfer characteristics with the surrounding coolant. These localized effects can influence the temperature of the nearby cladding surface and introduce uncertainties in interpreting the thermocouple data. Understanding the impact of thermocouples attached to the outer surface of the cladding is crucial for accurate data interpretation as well as its effect on the thermomechanical behavior of the cladding. This paper presents a novel methodology for simulating the impact of outer cladding thermocouples during transient testing of nuclear fuels. The simulation framework leverages the thermal-hydraulic capabilities of RELAP5-3D coupled to the BISON fuel performance code through the RELAPCouplingApp interface. The methodology is compared against Accident Tolerant Fuel Reactivity Initiated Accident-1-E experiment performed at Idaho National Laboratory. The results reveal approximately 100°C difference between thermocouple-altered temperature and virgin cladding surface. The model overpredicts surface rewet time due to conservative correlations.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Data Evaluation of Actinide Cross Sections: 240 Am, 241 Am, and 242 Am

Evaluations for neutron-induced reactions on the isotopes 240 Am, 241 Am, and 242 Am were per formed by optimizing reaction-model calculations based on existing experimental data from both direct measurements and surrogate reactions. The evaluated data were based on results obtained from the LLNL reaction codes YAHFC and FRESCO using up-to-date nuclear models; including an improvement upon some previous treatments for inelastic processes. Nuclear reaction codes generated data for all relevant reaction channels, which were then translated into Generalized Nu clear Data Structure and LLNL’s Evaluated Neutron Data Libraries (ENDL). These data libraries were tested for potential errors and applied to verification and validation (V&V) tests suites where applicable for comparisons.

42 ENGINEERING↗
DOE OSTI2017

PNNL Measurement Results for the 2016 Criticality Accident Dosimetry Exercise at the Nevada National Security Stite (IER-148)

The Pacific Northwest National Laboratory (PNNL) participated in a criticality accident dosimetry intercomparison exercise held at the Nevada National Security Site (NNSS) May 24-27, 2016. The exercise was administered by Lawrence Livermore National Laboratory (LLNL) and consisted of three exposures performed using the Godiva-IV critical assembly housed in the Device Assembly Facility (DAF) located on the NNSS site. The exercise allowed participants to test the ability of their nuclear accident dosimeters to meet the performance criteria in ANSI/HPS N13.3-2013, Dosimetry for Criticality Accidents and to obtain new measurement data for use in revising dose calculation methods and quick sort screening methods where appropriate. PNNL participated with new prototype Personal Nuclear Accident Dosimeter (PNAD) and Fixed Nuclear Accident Dosimeter (FNAD) designs as well as the existing historical PNAD design. The new prototype designs incorporate optically stimulated luminescence (OSL) dosimeters in place of thermoluminescence dosimeters (TLDs), among other design changes, while retaining the same set of activation foils historically used. The default dose calculation methodology established decades ago for use with activation foils in PNNL PNADs and FNADs was used to calculate neutron dose results for both the existing and prototype dosimeters tested in the exercise. The results indicate that the effective cross sections and/or dose conversion factors used historically need to be updated to accurately measure the operational quantities recommended for nuclear accident dosimetry in ANSI/HPS N13.3-2013 and to ensure PNAD and FNAD performance meets the ANSI/HPS N13.3-2013 performance criteria. The operational quantities recommended for nuclear accident dosimetry are personal absorbed dose, Dp(10), and ambient absorbed dose, D*(10).

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2023

Updated evaluation of potential ultralow Q -value β -decay candidates

“Ultralow” Q-value β decays are referred to as such due to their low decay energies of less than ≈1 keV. Such a low energy decay is possible when the parent nucleus decays to an excited state in the daughter, with an energy close to that of the Q value. These decays are of interest as potential new candidates for neutrino mass determination experiments and as a testing ground for studies of atomic interference effects in the nuclear decay process. In this paper, we provide an updated evaluation of atomic mass data and nuclear energy-level data to identify potential ultralow Q-value β decay candidates. For many of these candidates, more precise and accurate atomic mass data is needed to determine if the Q value of the potential ultralow decay branch is energetically allowed and in fact ultralow. Furthermore, the relevant precise atomic mass measurements can be achieved via Penning trap mass spectrometry.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2015

Concepts for Small-Scale Testing of Used Nuclear Fuel

This report documents a concept for a small-scale test involving between one and three Boiling Water Rector (BWR) high burnup (HBU) fuel assemblies. This test would be similar to the DOE funded High Burn-Up (HBU) Confirmatory Data Project to confirm the behavior of used high burn-up fuel under prototypic conditions, only on a smaller scale. The test concept proposed would collect data from fuel stored under prototypic dry storage conditions to mimic, as closely as possible, the conditions HBU UNF experiences during all stages of dry storage: loading, cask drying, inert gas backfilling, and transfer to an Independent Spent Fuel Storage Installation (ISFSI) for multi-year storage.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Code Benchmark of the HTTF Pressurized Conduction Cooldown Test Using SAM

The High Temperature Test Facility (HTTF) at Oregon State University is an integral system test facility to simulate postulated reactor transients of prismatic high-temperature gas-cooled reactors(HTGRs). A series of test campaigns was launched, providing abundant test data that could be used to benchmark reactor system analysis codes like the System Analysis Module (SAM). In this study, a SAM model of the facility is developed based on the two-dimensional (2D) ring model approach. All components including the ceramic matrix, graphite heaters, helium coolant channels, core barrel, upcomer, pressure vessel, and reactor cavity cooling system are modeled as concentric cylindrical rings. The model is used to simulate one of the benchmark problems-Pressurized Conduction Cooldown (PCC)-within the scope of the Organisation for Economic Co-operation and Development Nuclear Energy Agency International HTTF Benchmark. The simulations consist of two parts. In the first part, operating and boundary conditions as well as thermophysical properties of materials are specified for the benchmark problem. In this work, results from the first part will be used in code-to-code comparison. In the second part, the SAM model is used to simulate Test PG-27, which is the first PCC test carried out in the HTTF, with only two of the ten heater banks activated. The results in the second part are used for code-to-data comparison. Because the helium coolant flow rate is not measured in this facility, it is estimated using the input power and inlet/outlet coolant temperatures. Additionally, radial heat flow in the ceramic blocks is complicated by hundreds of cylindrical coolant channels and heater rods embedded in them. As such, it is necessary to deduce an effective thermal conductivity for the ceramic to analyze the core thermal behavior. SAM predictions of the helium coolant and ceramic temperatures are compared with test data measured in three equivalent sectors. Overall, the SAM results agree reasonably well with test data within the variation of data among the three sectors, which demonstrates SAM's capability in capturing transient effects in HTGR using the simplified 2D ring model.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Transient Testing of Nuclear Fuels Performed in the Original Operation of TREAT (2nd ed.)

This report provides summary information on experiments performed in the Transient Reactor Test Facility (TREAT) during its original operation from 1959-1994, and more specifically, before the TREAT core was upgraded at the end of the 1980s. The tests performed since about 1970 are summarized in considerable detail. Earlier tests are noted by their designated series and by some parameters that help to identify and distinguish each series from the others. The detailed summary descriptions are intended to help inform future researchers regarding the large, historical, empirical data base generated from that 35-year span during which hundreds of TREAT experiments were performed to investigate and demonstrate the transient behavior of a wide variety of nuclear fuels during severe off-normal and accident conditions -- using direct nuclear heating of the test samples. That empirical basis is foundational and valuable for guiding future qualification and safety evaluation of improved fuel designs and for future transient test planning and performance in TREAT. Since Edition 1 of this report, descriptions of the following test series have been added: D, E, EOS, F, H, J, R, RX, and S.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗