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At least 667 records · Page 37
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DOE OSTI2017

ATF-1 Power Histories

MCNP5 calculated ATF-1 power histories are documented in this Technical Evaluation (TEV). The current revision of this TEV (Revision 0), includes power histories through Advanced Test Reactor (ATR) Cycle 162A. The Cycle 162A power histories are based off projection calculations while all other cycles are based off as-run (core follow) simulations. This TEV will serve as the single source for ATF-1 experiment power history information and will be updated after each cycle of operation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Current status of post-irradiation examination for the ATF-1 irradiation

A series of irradiations, referred to as the ATF series, is currently underway to evaluate the feasibility of different Accident Tolerant Fuels for commercial light water reactors. For initial screening, concepts consisting of an uranium bearing phase and cladding were irradiated in stainless steel capsules. These tests provide screening data among concepts to ensure viability of a fuel form going forward and can provide some initial data on fuel performance particularly information about fuel-cladding interaction and fuel behavior at operating temperatures and fission densities. Screening concepts are under irradiation or planned for irradiation at the Advanced Test Reactor of Idaho National Laboratory for a number of industrial partners (AREVA, GE, Westinghouse) and national laboratories (Los Alamos National Laboratory, Oak Ridge National Laboratory). Baseline PIE will be performed on all irradiated rodlets and will consist of visual examinations, neutron radiography, gamma spectrometry, dimensional inspections, fission gas analysis, optical microscopy, and chemical analysis. Additional evaluations may also be performed. The first capsules have arrived at the hot cell for initial post-irradiation examination. Results from initial examinations are presented along with example results from equivalent exams on similar fuel. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Recombination of Hydrogen in the Iodine Reactors

The H-Canyon facility is currently dissolving spent nuclear fuel, including Material Test Reactor (MTR) and High Flux Isotope Reactor (HFIR) fuel. Dissolution of aluminum spent nuclear fuel produces hydrogen and other NOx gases. A theory long held by H-Canyon Engineering and Facility Technical Advisors that the silver nitrate-coated berl saddles, present in the iodine reactor, used for off-gas treatment were acting as a catalytic hydrogen recombiner was captured during an external review of the Accelerated Basin Deinventory program. Results from a separate SRNL project using a Raman Spectrometer to monitor the offgas from the dissolution process indicated a lack of the expected hydrogen in the off-gas stream. To test this hypothesis, a laboratory scale iodine reactor was assembled and filled with silver nitrate-coated berl saddles. Testing with this laboratory scale reactor confirmed the recombination of hydrogen when a simulated dissolver off-gas was passed through the reactor containing silver nitrate-coated berl saddles at 173-188 °C. Control experiments performed with uncoated berl saddles resulted in no change to the hydrogen concentration after passing through the reactor. The residence time of the gas in the reactor was varied to determine if residence time had an impact on the amount of hydrogen recombination occurring. Results from these experiments indicated that at the shortest residence times tested (~5 seconds) recombination of the hydrogen still exceeded 90%; however, the percent recombination did increase at longer residence times, reaching 97.9% or greater for residence times over 10 seconds. Finally, testing performed with a simplified off-gas composition containing only hydrogen and air gave similar results, indicating that the presence of NO and N 2 O gases in the stream do not play a role in the recombination reaction.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

On-line monitoring of I and C transmitters and sensors for calibration verification and response time testing was successfully implemented at ATR - 326

The Advanced Test Reactor (ATR) has always had a comprehensive procedure to verify the performance of its critical transmitters and sensors, including RTDs, and pressure, level, and flow transmitters. These transmitters and sensors have been periodically tested for response time and calibration verification to ensure accuracy. With implementation of on-line monitoring techniques at ATR, the calibration verification and response time testing of these transmitters and sensors are verified remotely, automatically, hands off, include more portions of the system, and can be performed at almost any time during process operations. The work was done under a DOE funded SBIR project carried out by AMS. As a result, ATR is now able to save the manpower that has been spent over the years on manual calibration verification and response time testing of its temperature and pressure sensors and refocus those resources towards more equipment reliability needs. More importantly, implementation of OLM will help enhance the overall availability, safety, and efficiency. Together with equipment reliability programs of ATR, the integration of OLM will also help with I and C aging management goals of the Department of Energy and long-time operation of ATR. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Initial Comparison of Direct and Legacy Modeling Approaches for Radial Core Expansion Analysis

Radial core expansion in sodium-cooled fast reactors provides an important reactivity feedback effect. As the reactor power increases due to normal start up conditions or accident scenarios, the core and surrounding materials heat up, causing both grid plate expansion and bowing of the assembly ducts. When the core restraint system is designed correctly, the resulting structural deformations introduce negative reactivity which decreases the reactor power. Historically, an indirect procedure has been used to estimate the reactivity feedback due to structural deformation which relies upon perturbation theory and coupling legacy physics codes with limited geometry capabilities. With advancements in modeling and simulation, radial core expansion phenomena can now be modeled directly, providing an assessment of the accuracy of the reactivity feedback coefficients generated by indirect legacy methods. Recently a new capability was added to the PROTEUS-SN unstructured geometry neutron transport solver to analyze deformed meshes quickly and directly. By supplying the deformed mesh in addition to the base configuration input files, PROTEUS-SN automatically processes material adjustments including calculation of region densities to conserve mass, calculation of isotopic densities according to material models (for example, sodium density as a function of temperature), and subsequent re-homogenization of materials. To verify the new capability of directly simulating deformed meshes, PROTEUS-SN was used to compute reactivity feedback for a series of contrived yet representative deformed configurations for the Advanced Burner Test Reactor design. The indirect legacy procedure was also performed to generate reactivity feedback coefficients for the same deformed configurations. Interestingly, the legacy procedure consistently overestimated reactivity feedbacks by 35% compared to direct simulations by PROTEUS-SN. This overestimation indicates that the legacy procedures are in fact not conservative and could be overestimating reactivity feedback effects that are closely tied to reactor safety. We conclude that there is indeed value in performing direct simulation of deformed meshes despite the increased computational expense. PROTEUS-SN is already part of the SHARP multi-physics toolkit where both thermal hydraulics and structural mechanical feedback modeling can be applied but this is the first comparison of direct simulation to legacy techniques for radial core expansion.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Submission of FeCrAl Feedstock for Support of AFC ATR-2 Irradiations

The Advanced Test Reactor (ATR) is currently being used to test accident tolerant fuel (ATF) forms destined for commercial nuclear power plant deployment. One irradiation program using the ATR for ATF concepts, Accident Tolerant Fuel-2 (ATF-2), is a water loop irradiation test using miniaturized fuel pins as test articles. This complicated testing configuration requires a series of pre-test experiments and verification including a flowing loop autoclave test and a sensor qualification test (SQT) prior to full test train deployment within the ATR. In support of the ATF-2 irradiation program, Oak Ridge National Laboratory (ORNL) has supplied two different Generation II FeCrAl alloys in rod stock form to Idaho National Laboratory (INL). These rods will be machined into dummy pins for deployment in the autoclave test and SQT. Post-test analysis of the dummy pins will provide initial insight into the performance of Generation II FeCrAl alloys in the ATF-2 irradiation experiment as well as within a commercial nuclear reactor.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Coupling the System Analysis Module with SAS4A/SASSYS-1

SAS4A/SASSYS-1 is a simulation tool used to perform deterministic analysis of anticipated events as well as design basis and beyond design basis accidents for advanced reactors, with an emphasis on sodium fast reactors. SAS4A/SASSYS-1 has been under development and in active use for nearly forty-five years, and is currently maintained by the U.S. Department of Energy under the Office of Advanced Reactor Technology. Although SAS4A/SASSYS-1 contains a very capable primary and intermediate system modeling component, PRIMAR-4, it also has some shortcomings: outdated data management and code structure makes extension of the PRIMAR-4 module somewhat difficult. The user input format for PRIMAR-4 also limits the number of volumes and segments that can be used to describe a given system. The System Analysis Module (SAM) is a fairly new code development effort being carried out under the U.S. DOE Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. SAM is being developed with advanced physical models, numerical methods, and software engineering practices; however, it is currently somewhat limited in the system components and phenomena that can be represented. For example, component models for electromagnetic pumps and multi-layer stratified volumes have not yet been developed. Nor is there support for a balance of plant model. Similarly, system-level phenomena such as control-rod driveline expansion and vessel elongation are not represented. This report documents fiscal year 2016 work that was carried out to couple the transient safety analysis capabilities of SAS4A/SASSYS-1 with the system modeling capabilities of SAM under the joint support of the ART and NEAMS programs. The coupling effort was successful and is demonstrated by evaluating an unprotected loss of flow transient for the Advanced Burner Test Reactor (ABTR) design. There are differences between the stand-alone SAS4A/SASSYS-1 simulations and the coupled SAS/SAM simulations, but these are mainly attributed to the limited maturity of the SAM development effort. The severe accident modeling capabilities in SAS4A/SASSYS-1 (sodium boiling, fuel melting and relocation) will continue to play a vital role for a long time. Therefore, the SAS4A/SASSYS-1 modernization effort should remain a high priority task under the ART program to ensure continued participation in domestic and international SFR safety collaborations and design optimizations. On the other hand, SAM provides an advanced system analysis tool, with improved numerical solution schemes, data management, code flexibility, and accuracy. SAM is still in early stages of development and will require continued support from NEAMS to fulfill its potential and to mature into a production tool for advanced reactor safety analysis. The effort to couple SAS4A/SASSYS-1 and SAM is the first step on the integration of these modeling capabilities.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Testing of graphite thermal neutron scattering law data in support of treat neutronic analysis

The Transient Reactor Test Facility (TREAT) has historically provided the capability for transient fuel performance and safety testing in the United States. The operation of this facility was suspended in 1994. However, in recent years preparations for the restart of TREAT were initiated. TREAT is a graphite moderated and air cooled reactor. Its core is composed of a 19 x 19 array of fuel assemblies that contain UO{sub 2} (HEU) fuel particles dispersed in a graphite matrix, along with reflector assemblies filled with graphite only. Each fuel assembly is 8 feet tall with a fueled region that is 4 feet tall and upper and lower regions of graphite that are 2 feet tall. In addition, a 2 feet thick radial graphite reflector region is present. Finally, a 5 feet thick concrete biological shield surrounds the core. Previous work showed that the graphite thermal scattering cross sections can introduce variations exceeding 2000 pcm in the predicted eigenvalue of the TREAT core. Recognizing that the graphite in the assembly (and thereby the core) is a porous mixture of crystalline and amorphous carbon (with a lower density in comparison to ideal crystalline graphite), it becomes clear that TREAT specific thermal neutron scattering cross section libraries will be needed (at all temperatures of interest) for use within neutronic codes seeking to simulate TREAT. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

The deformation and burst behavior of Zircaloy-4 cladding tubes with hydride rim features subject to internal pressure loads

This study is motivated by the desire to conduct a separate effects test replicating specific aspects of an in-pile pulse reactor test on high exposure light water reactor (LWR) fuel rods. The principal purpose of the tests was to determine whether temperature and or hydrogen concentration thresholds exist in assessing the vulnerability of a stress relieved zirconium alloy cladding tubes to low strain ruptures. A secondary objective was to determine whether observations of the cladding stress during deformation could aid in the understanding of the crack formation and propagation in cladding tubes with hydride rim features. Here, the study utilized artificially hydrided Zircaloy-4 cladding tubes and an isothermal, internal pressure test that was controlled with an in-situ strain measurement. The test results show a clear transition in failure vulnerability between room temperature and 150 °C. Below 150 °C, cladding tubes rupture at very low uniform plastic strain—even at relatively modest cladding hydrogen levels. At and above 150 °C, cladding tubes only rupture at low uniform plastic strain if the cladding hydrogen content is greater than 500 ppm and usually experience at least some modest amount of plastic deformation prior to rupture. Post-test metallographs from non-ruptured cladding tubes show numerous primary cracks forming in the hydride rim and arresting at the metal/hydride interface. This observation combined with the decrease in bulk yielding of hydrided cladding tubes when compared to fresh non-hydrided tubes leads to the conclusion that the fracture stress of the hydride rims is related to and bounded by the yield stress of the zirconium metal. Some evidence is presented for a change in crack propagation mechanism from one of void formation on hydride plates and linkage at temperatures below 150 °C to more classical ductile shearing beneath the primary crack tip at temperatures above 150 °C. This apparent change in crack propagation mechanism may help explain the observed temperature and hydrogen concentration thresholds.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Derivation of the Point-Kinetics Parameters for the Coupled Reactors

The design of coupled reactors, which provide two distinct spectral zones (fast and thermal), has recently drawn attention because it can provide configurations that could not be met in current thermal neutron test reactors and thus allows versatile irradiation tasks. A coupled reactor core is characterized by the neutron multiplicity of each zone as well as by the coupling between the two zones. It is desired to evaluate the integral point kinetics parameters that characterize the neutron physics of the coupled system. This summary provides a derivation of the equations for evaluating these point kinetics parameters. Following the strong and the weak form of the multigroup diffusion equation, the weak forms for the partial forward and adjoint solutions, each for a particular active region of the reactor core, are derived. These partial solutions are then used to collapse the weak form into the coupled-reactor point kinetics equations with the approximation of equilibrium. It is noted that these parameters are not meant for transient analysis because of the equilibrium approximation. They are rather used for characterizing the coupled system. The derived reactivity equations are then compared with those in Ref. 1. The equations are implemented in Rattlesnake, the INL MOOSE-based radiation transport application for multi-physics simulations. Numerical results are presented for a simple two-group problem. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

HIGH BURNUP FUEL-COOLANT INTERACTION ANALYSIS SUPPORTING FUEL SAFETY TESTING AT IDAHO NATIONAL LABORATORY

In the near future, experiments on HBu fuel under loss-of-coolant accident (LOCA) and reactivity-initiated accident (RIA) conditions will be performed within the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL). These experiments will be performed using the Transient Water Irradiation System for TREAT (TWIST) experiment vehicle. To support these experiments, analysis of fuel-coolant interaction (FCI) energetics is underway. This paper discusses FCIs in the context of light water reactor (LWR) safety, differentiating between the severe accident focus of commercial reactors and experimental RIA test programs where FCIs have occurred. However, it is highlighted that as the nuclear industry aims for increased burnup limits, the FCI events observed in RIA test programs may become relevant to commercial LWR safety analysis. The paper then presents developments to the UW-FCI computer program to enable simulation of FCIs initiated by solid fuel particles dispersing into the coolant during RIAs.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

BISON fuel performance modeling optimization for experiment X447 and X447A using axial swelling and cladding strain measurements

With the recent need to qualify new reactor designs such as the Versatile Test Reactor (VTR), fuel performance calculations need to be performed to determine safety criteria of the proposed designs. In order to validate the fuel performance results obtained by a fuel performance code, BISON, for new reactor designs, legacy fuel from EBR-II and FFTF MFF with Post -Irradiation Examination (PIE) data need to be used as validation cases to benchmark models. Here in this work, BISON has been paired with the Fuels Irradiation & Physics Database (FIPD) and IFR Materials Information System (IMIS) to supply PIE data for comparison with simulations of EBR-II experiments X447/X447A. X447/X447A were assessed by implementing models for Fuel Cladding Chemical Interaction (FCCI) within BISON and optimizing the friction coefficient between the fuel surface and the cladding, the anisotropic swelling factor, and the HT9 first thermal creep scalar (which scales the first term in the HT9 creep equation) to best match the PIE axial fuel swelling height and cladding profilometry for all pins in X447/X447A. The optimal values were found using a generic algorithm developed to select different values for the three parameters until end criteria was met and error couldn’t be reduced further. The BISON-simulated cladding profilometry was evaluated using Standard Error of the Estimate (SEE) to account for the profile shape of the cladding profilometry. Optimal values for the friction coefficient, anisotropic fuel swelling factor, and HT9 first thermal creep scalar were found to best fit the BISON simulation results to the PIE measurements found in IMIS and FIPD. Improvements to current models are suggested to account for the underprediction of fuel swelling at low burnups and the overprediction of fuel swelling at higher burnups observed for the axial fuel swelling height. Although two pins in EBR-II X447/X447A (DP70 and DP75) were known to fail due to FCCI, none of the pins simulated in BISON reached a cumulative damage fraction (CDF) above 0.008 with FCCI correlations coupled in the BISON simulations. The error estimate generated for all pins in X447/X447A using optimal values was 209 µm, which is deemed acceptable.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

SSHAC Level 1 Probabilistic Seismic Hazard Analysis for the Idaho National Laboratory

A Probabilistic Seismic Hazard Analysis (PSHA) was completed for the Materials and Fuels Complex (MFC), Naval Reactors Facility (NRF), and the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) (Figure 1-1). The PSHA followed the approaches and procedures appropriate for a Study Level 1 provided in the guidance advanced by the Senior Seismic Hazard Analysis Committee (SSHAC) in U.S. Nuclear Regulatory Commission (NRC) NUREG/CR-6372 and NUREG-2117 (NRC, 1997; 2012a). The SSHAC Level 1 PSHAs for MFC and ATR were conducted as part of the Seismic Risk Assessment (SRA) project (INL Project number 31287) to develop and apply a new-risk informed methodology, respectively. The SSHAC Level 1 PSHA was conducted for NRF to provide guidance on the potential use of a design margin above rock hazard levels. The SRA project is developing a new risk-informed methodology that will provide a systematic approach for evaluating the need for an update of an existing PSHA. The new methodology proposes criteria to be employed at specific analysis, decision, or comparison points in its evaluation process. The first four of seven criteria address changes in inputs and results of the PSHA and are given in U.S. Department of Energy (DOE) Standard, DOE-STD-1020-2012 (DOE, 2012a) and American National Standards Institute/American Nuclear Society (ANSI/ANS) 2.29 (ANS, 2008a). The last three criteria address evaluation of quantitative hazard and risk-focused information of an existing nuclear facility. The seven criteria and decision points are applied to Seismic Design Category (SDC) 3, 4, and 5, which are defined in American Society of Civil Engineers/Structural Engineers Institute (ASCE/SEI) 43-05 (ASCE, 2005). The application of the criteria and decision points could lead to an update or could determine that such update is not necessary.

58 GEOSCIENCES↗
DOE OSTI2020

As-Run Neutronics Evaluation for the CSM-10584 Experiment in the ATR

This engineering calculations and analysis report (ECAR) documents the results of the Advanced Test Reactor (ATR) detailed Monte Carlo N-Particle (MCNP) code full-core model as-run physics analysis performed to support the Colorado School of Mines (CSM) experiment in the B-5 position.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Startup Physics Testing of Advanced Reactors

This report documents the startup physics testing from initial fuel loading through ascension to full power for past advanced reactor startup physics testing programs. The review includes an assessment of what nuclear physics data was measured, why this data was measured, how was the measurement made, and the agreement with predictive reactor performance calculations of that time. The purpose of this review is to establish historical precedence for test inclusion for future advanced reactors planned for demonstration at the National Reactor Innovation Center (NRIC). The historical review includes reactor designs considered to be significantly different from current light-water reactor designs, or use simplified, inherent, passive, or other innovative means to accomplish their safety functions.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

DCIS integration tests for Lungmen nuclear power plant - 370

The factory acceptance tests (FAT) for distributed control and information systems (DCIS) of Lungmen nuclear power plant (LMNPP) Units 1 and 2 were separately performed by individual suppliers for their supplied scopes. The FAT also included overlap interface tests between the separately supplied scopes, but did not include an overall integration test. There were concerned by the regulator and the public over the lack of an integration test, which was believed to be indispensable to verify the correctness and completeness of a digital system of such large scale ever applied by Taiwan Power Company (Taipower). To address these concerns and enhance confidence in the DCIS, the DCIS integration tests, including reactor water level integration test, reactor pressure integration test, integration test for turbine generator and reactor protection system, integration test for proper plant response to safety and non-safety network equipment failures, and DCIS dynamic integration test, were performed at site during Unit 1| site tests. It is to be noted that these tests were beyond the post construction tests, the pre-operational tests, and the start-up tests, which are the normal industry practice and formally required by the regulator. A common goal for these integration tests is to verify the capability of the DCIS to transmit signal/data properly across various suppliers' equipment to actuate the final equipment. The methods and principle, the test purpose/test items/acceptance criteria and the test results for these tests are described. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗