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DOE OSTI1992

Nuclear Safety [Vol. 33, No. 4, October-December 1992]

Nuclear Safety is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 499 Technical Note: Nuclear Regulatory Commission’s Examination Question Bank System, C. L Jensen and N. L. Skinner; 506 Report on the Specialists’ Meeting on Passive and Active Safety Features of Liquid-Metal Fast Breeder Reactors Organized by the International Atomic Energy Agency at Oarai Engineering Centre of Power Reactor and Nuclear Development Corporation, Japan, November 5-7,1991, S. R. Paranjpe; ACCIDENT ANALYSIS: 514 A Review of Hydrogen Production During Melt/Water Interaction in LWRs, D. F. Fletcher, B. D. Turland, and S. P. A. Lawrence; 535 A Review of the U 3 0 8 -Al Reaction as a Potential Heat Source in Research and Test Reactor Accidents, J. L. Snelgrove; CONTROL AND INSTRUMENTATION: 543 Multilevel Interfaces for Power Plant Control Rooms II: A Preliminary Design Space, K. J. Vicente; DESIGN FEATURES: 549 Twenty-Second DOE/NRC Nuclear Air Cleaning and Treatment Conference, R. R. Bellamy, D. W. Moeller, and M. W. First; 562 Book Review: Safety Assessment of Unit 5 (WWER-440/W-213) of the Greifswald Nuclear Power Station, M. H. Fontana; WASTE AND SPENT FUEL MANAGEMENT: 576 Activities Related to Waste and Spent Fuel Management, M. D. Muhlheim and E. G. Silver; OPERATING EXPERIENCES: 586 Analysis of Corrective Actions Applied to Nuclear Power Plant Operations, A. E. Cross; 593 Reactor Shutdown Experience, Compiled by J. W. Cletcher; 596 Operating U.S. Power Reactors, Compiled by M. D. Muhlheim and E. G. Silver; 609 Non-Commercial Power Operating Experience; 614 World Nuclear Performance, Compiled by M. D. Muhlheim; RECENT DEVELOPMENTS: 623 General Administrative Activities, Compiled by M. D. Muhlheim and E. G. Silver; 635 Reports, Standards, and Safety Guides, D. S. Queener; 639 Proposed Rule Changes as of June 30, 1992; ANNOUNCEMENTS: 513 Fourth International Conference on Nuclear Fuel Reprocessing and Waste Management RECOD '94; 585 International Topical Meeting on Advanced Reactors Safety; 622 International Workshop on Implementation of ALARA at Nuclear Power Plants; 648 Short Course on Nuclear Criticality Safety; 648 8th Topical Meeting on Problems in Reactor Physics MEPhi-VOLGA-93; 643 The Authors; 646 Reviewers of Nuclear Safety, Vol. 33

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

AGR-5/6/7 Experiment Monitoring and Simulation Progress

Advanced Gas Reactor (AGR)--5/6/7 is the last of a series of AGR experiments conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) in support of development and qualification of tri-structural isotropic (TRISO) low-enriched fuel for use in high temperature gas-cooled reactors. The test train contains five separate capsules that are independently controlled and monitored. Each capsule contains multiple 12.51 mm long compacts filled with low-enriched uranium carbide/oxide (UCO) TRISO fuel particles. The objectives of the AGR-5/6/7 experiment are to: 1. Irradiate reference-design fuel particles to support fuel qualification. 2. Establish operating margins for the fuel beyond normal operating conditions. 3. Provide irradiated fuel performance data and irradiated fuel samples for post-irradiation examination (PIE) and safety testing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

AGR-5/6/7 Experiment Monitoring and Simulation Progress

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Evaluation of the NSUF Reactor Activation and Damage (RAD) Calculator damage component

The Nuclear Science User Facilities (NSUF) provide researchers access to nuclear research facilities and expertise all over the United States and at the Belgian Center for Nuclear Research (SCK/CEN). One special capability of the NSUF is providing access to nuclear research and test reactors often used for conducting material testing experiments. These services are provided at no cost to the researcher through a competitive, peer-reviewed process. With so many proposals being submitted simultaneously, an NSUF information management resource called the Combined Materials Experiment Toolkit (CoMET) was implemented to speed up the review process and give users access to the information and tools necessary for designing meaningful experiments and writing the best proposals possible. One of CoMET’s tools is the Reactor Activation and Damage (RAD) Calculator. For those with little or no background in it, neutron irradiation in research and test reactors can be a challenging subject to include in a proposal. That is why the RAD Calculator was created to assist users with providing estimates of irradiation and post-irradiation conditions, enabling them to select the best neutron irradiation and post-irradiation examination facilities for reaching their experiment goals. The RAD Calculator is a general scoping tool that estimates radiation damage in terms of DPA, in addition to the resulting radioactivity of the material post-irradiation. This report evaluates the neutron damage component of the RAD Calculator. The evaluation of the activation component of the calculator is covered in report INL/EXT-20-58080.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2019

AGR-5/6/7 Experiment Monitoring and Simulation Progress

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2019

Fuel Plate Surface Temperature in a Fully-Uncovered Fuel Assembly after a Loss-of-Coolant Accident in an MTR-Type Research Reactor

A methodology and model that were developed in the 1960s for calculating the peak fuel temperature in an MTR fuel assembly in a reactor core that has been fully-uncovered in a hypothetical loss-of-coolant accident (LOCA) are revived and updated in this report. The semi-empirical one-node model is based on experiments on fuel elements for the Oak Ridge Research Reactor (ORR) and the Low Intensity Testing Reactor (LITR) that were done in the 1950s and 1960s at the Oak Ridge National Laboratory. The peak plate temperature in 18 LOCA tests in the ORR and two of 24 tests in the LITR, calculated using the model, are compared with experiments. The original model was used in LOCA analyses for the Omega West Reactor (OWR) at Los Alamos National Laboratory with HEU fuel and for three cores of the University of Virginia Reactor (UVAR) with both HEU and LEU fuels. The LOCA analyses for UVAR using this model were accepted for licensing purposes by the US Nuclear Regulatory Commission for an assembly containing 12 HEU fuel plates in 1970, an assembly containing 18 HEU fuel plates in 1984, and an assembly containing 22 LEU silicide fuel plates in 1994. The analyses done for the OWR and UVAR are repeated here and compared with data in the original reports in order to establish the validity of the model revived in this report. The model is then used in a subsequent LOCA analysis for the HEU and LEU cores of the RPI MTR-type reactor in Portugal. It can also be applied to other MTR-type reactors with designs similar to the LITR.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2016

The ATWS analysis of the HTR-10GT core induced by the earthquake - 18489

The 10 MW High Temperature Gas Cooled Test Reactor (HTR-10) has been built in Institute of Nuclear and New Energy Technology (INET) and has been operating successfully since the beginning of 2003. The core outlet temperature of HTR-10 is 700 deg. C. To verify the technology of gas-turbine direct cycle, at first INET had a plan to increase its core outlet temperature to 750 deg. C and use a helium gas turbine instead of the steam generator (then the reactor is called HTR-10GT). Though HTR-10 has good intrinsic safety, the design basic accidents and beyond design basis accidents of HTR-10GT must be analyzed according to China s nuclear regulations due to changed operation parameters. THERMIX code system is used to study the Anticipated Transient Without Scram (ATWS) caused by an earthquake. The gravity acceleration of the earthquake will make the porosity of the pebble bed smaller, which will thus induce a positive reactivity. The packing fraction of the HTR-10GT core will increase to 0.64 from 0.61, while a reactivity of 1.24% will be inserted into an initial core and 0.788% for an equilibrium core. Then the reactor power increased and the temperature of the core increased. When the neutron flux of power measuring range exceeded 123% and the reactor period becomes less than 20 seconds, the reactor should scram. It was supposed that all the control rods in the reflectors had been blocked and the reactor could not scram. Thus the accident went on and the core temperature and the system pressure increased but the reactor shutdown at last because of its natural negative temperature reactivity feedback mechanism. The residual heat would be removed out of the core by the cavity cooling system. During the accident sequence the maximum fuel temperature was 1223 deg. C. If a temperature coefficient error of 20% was considered during analyzing, the peak temperature of the fuel during the ATWS will reach 1266 deg. C. It was a little higher than 1230 deg. C --the fuel temperature limitation of HTR-10. Now the sphere fuel used in HTR-10GT will also be used in HTR-PM and the temperature limitation has been raised to 1620 deg. C, so the HTR-10GT is safe during the ATWS caused by an earthquake. The paper also compares the analysis result of HTR10-GT to those of HTR-10. The results shows that the HTR-10GT is still safe during the accident though its operating temperature is higher than HTR-10. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2023

Resumption of water capsule reactivity-initiated accident testing at TREAT

A series of integral reactivity-initiated accident (RIA) commissioning experiments were completed in a new static water capsule in the Transient Reactor Test Facility (TREAT), marking the first such tests in the U.S. in more than 40 years. The test campaign included a verification test followed by five tests in the Static Environment Rodlet Transient Test Apparatus (SERTTA) capsule. The capsule initial conditions varied from room temperature and pressure up to 200°C and 2.5 MPa, with energy depositions varying between ~500–1100 J/gUO 2 . The series of tests allowed for a number of instrumentation qualifications and demonstrations, including cladding thermometry, rodlet plenum pressure, cladding elongation, and an electro-impedance boiling detector. Post transient examinations such as gamma emission spectroscopy, profilometry, and microscopy were performed to document the end state of the fuel rods. Furthermore, the results from the experiments show that the fuel rodlets behaved very similar to historical tests under similar energy depositions. This paper documents the design of the capsule and highlights some results from the commissioning tests and post-transient examination.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Device for steam cladding oxidation testing at TREAT

To compare the chemical degradation of conventional zirconium alloy (Zry) cladding to advance silicon carbide (SiC) cladding in a post loss of coolant accident (LOCA) environment, new nuclear testing capabilities are necessary. The Transient Reactor Test (TREAT) Facility at Idaho National Laboratory (INL) has matured its transient fuel testing capabilities since its 2017 restart. The most recent experiment architecture is the Transient Water Irradiation System in TREAT (TWIST), which is designed to support qualification of accident tolerant fuels in light water reactors. INL has designed and analyzed a natural circulation steam flow modification for TWIST to produce prototypic conditions of cladding oxidation. The in-situ device will be electrically heated to drive natural circulation. Moreover, the SiC cladding requires heating above 1700 °C to observe failure, thus internal prototypic nuclear heating with radiation effects will be used. Thermal hydraulic analysis with RELAP5-3D (Reactor Excursion and Leak Analysis Program) estimated steam fluxes greater than 50 mg cm −2 s −1 can be achieved. These fluxes are adequate to test Zry cladding according to draft regulatory guides and to test SiC cladding according to past experiments.

Oxidation↗

DOE OSTI2023

Monte Carlo Analysis of Coolant Stream Impurity Gamma Emissions in Gas-Cooled Fast Reactors

It is well established that a rapid increase in the concentration of fission products in the reactor coolant stream can serve as an early indication of fuel failure. We use Monte Carlo simulations to investigate the feasibility of using several gamma detectors as diagnostic equipment to monitor the presence of major fission product isotopes in high-temperature gas-cooled reactor (HTGR) coolant streams for early detection of fuel failure and therefore the prevention of fuel failure conditions. Herein we model the response of high-purity germanium (HPGe), CdZnTe, NaI(Tl), and LaBr 3 (Ce) detectors of typical commercial sizes to the gamma emissions from nuclides expected to be found within the coolant stream of the Versatile Test Reactor (VTR) under development by the U.S. Department of Energy. The results indicate that for the 233- and 250-keV gamma rays from 133 Xe and 135 Xe, respectively, the 3σ detection criterion is met in under 1 min using a single HPGe detector. Changes in other spectral lines associated with Xe nuclides are detected within 1 h regardless of the choice of detector.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2016

Advanced Reactor Fuels Irradiation Experiment Design Objectives

This report summarizes the objectives of the current irradiation testing activities being undertaken by the Advanced Fuels Campaign relative to supporting the development and demonstration of innovative design features for metallic fuels in order to realize reliable performance to ultra-high burnups. The AFC-3 and AFC-4 test series are nearing completion; the experiments in this test series that have been completed or are in progress are reviewed and the objectives and test matrices for the final experiments in these two series are defined. The objectives, testing strategy, and test parameters associated with a future AFC test series, AFC-5, are documented. Finally, the future intersections and/or synergies of the AFC irradiation testing program with those of the TREAT transient testing program, emerging needs of proposed Versatile Test Reactor concepts, and the Joint Fuel Cycle Study program’s Integrated Recycle Test are discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Nuclear Requalification of the New ATR Core

The Advanced Test Reactor (ATR) creates a unique high neutron flux materials testing environment that subsequently causes neutron embrittlement damage to its beryllium reflector eventually requiring a replacement during a reactor Outage. Concurrently, a piece-for-piece core replacement is conducted to maintain reactor functionality called a core internals changeout (CIC). Once changeout is complete, ATR Reactor Engineering utilizes previous CIC data to model and predict criticality scenarios for the new unirradiated core. However, CIC data has been inconsistently reported over the years making it difficult to find data trends that could be useful for the modeling and prediction process. So nuclear requalification testing is conducted using different flux and irradiation tests to accurately verify the models and to recalibrate nuclear monitoring instrumentation in the ATR core allowing for safe re-start of normal operations.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2015

Uncertainty Quantification of Calculated Temperatures for the U.S. Capsules in the AGR-2 Experiment

A series of Advanced Gas Reactor (AGR) irradiation experiments are being conducted within the Advanced Reactor Technology (ART) Fuel Development and Qualification Program. The main objectives of the fuel experimental campaign are to provide the necessary data on fuel performance to support fuel process development, qualify a fuel design and fabrication process for normal operation and accident conditions, and support development and validation of fuel performance and fission product transport models and codes (PLN-3636). The AGR-2 test was inserted in the B-12 position in the Advanced Test Reactor (ATR) core at Idaho National Laboratory (INL) in June 2010 and successfully completed irradiation in October 2013, resulting in irradiation of the TRISO fuel for 559.2 effective full power days (EFPDs) during approximately 3.3 calendar years. The AGR-2 data, including the irradiation data and calculated results, were qualified and stored in the Nuclear Data Management and Analysis System (NDMAS) (Pham and Einerson 2014). To support the U.S. TRISO fuel performance assessment and to provide data for validation of fuel performance and fission product transport models and codes, the daily as-run thermal analysis has been performed separately on each of four AGR-2 U.S. capsules for the entire irradiation as discussed in (Hawkes 2014). The ABAQUS code’s finite element-based thermal model predicts the daily average volume-average fuel temperature and peak fuel temperature in each capsule. This thermal model involves complex physical mechanisms (e.g., graphite holder and fuel compact shrinkage) and properties (e.g., conductivity and density). Therefore, the thermal model predictions are affected by uncertainty in input parameters and by incomplete knowledge of the underlying physics leading to modeling assumptions. Therefore, alongside with the deterministic predictions from a set of input thermal conditions, information about prediction uncertainty is instrumental for the ART program decision-making. Well defined and reduced uncertainty in model predictions helps increase the quality of and confidence in the AGR technical findings.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Initial exploration of a novel transient arrest system involving fuel heating

A preliminary analysis on a novel accident response system to diminish the severity of super- critical transients was conducted. The novel accident response system, called the instant shock arrest system, involves using electricity to heat the nuclear fuel at the onset of a large accidental reactivity insertion. This system is specifically designed for reactors with metallic fuel, such that the fuel is capable of conducting electricity, and being resistively heated. A reactor dynamics model of the advanced test reactor was created using the point kinetics equations and a linear reactivity feedback model to simulate how the system would effect the maximum fuel temperatures experienced during the transient. Transients with the instant shock arrest system were compared to those without it. It was found that the instant shock arrest system initially heated the fuel more than the unaffected transient but the negative reactivity inserted from such heating was enough to lower the maximum fuel temperature experienced during the transient. After simulating six different accident scenarios with reactivity insertions ranging from 0.5 to 1.3 dollar, it was found that an optimal system response could reduce peak fuel temperatures during the transient by 3.5% to 5%. Furthermore, discussion was given on how the optimal system response could be obtained using relatively simple numerical optimization algorithms due to the smoothness of the optimization problem. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

Neutron and Ion Irradiation Studies on Advanced Steels Using the Nuclear Science User Facilities

For nuclear reactor structural materials, irradiation is capable of inducing a series of microstructural modifications, such as dislocation loops/network, selective elemental segregation and precipitation, voids and helium bubbles, leading to the degradation in structural materials' performance through irradiation hardening and embrittlement, swelling, irradiation creep, etc. Since radiation poses challenges for structural applications in nuclear reactors in multiple aspects and the underlying mechanisms are often interconnected, radiation effect studies need well-designed test matrix and proper characterization techniques. The Nuclear Science User Facilities (NSUF) offer extensive capabilities for studying radiation effects in nuclear fuels and structural materials. This research covers both neutron and ion irradiation studies carried out using a variety of instruments through the NSUF. The neutron irradiation ferritic ODS steel MA957 was accomplished utilizing the neutron irradiation capability of the Advanced Test Reactor (ATR) and the post-irradiation examination (PIE) capabilities of the Center for Advanced Energy Studies (CAES). The neutron-irradiated specimens were characterized using transmission electron microscopy (TEM) and atom probe tomography (APT). The in situ TEM ion irradiation study on austenitic stainless steel (SS) 316H was completed using the IVEM-Tandem facility at Argonne National Laboratory.

36 MATERIALS SCIENCE↗

DOE OSTI2019

MCNP Analysis fo the Aqua-SETH Experiment in TREAT

The versatile Transient Reactor Test (TREAT) Facility’s capability to deliver customized transient nuclear power excursions gives researchers and developers a multitude of specialized testing options. In conjunction with reactorprescribed delivery of neutron flux, experiment design requires specimen packaging to meet TREAT safety containment requirements and experimenter test objectives. These objectives may range from the very simple separate effects tests (generally dry capsule) to the more complex integral scale testing wherein prototypic boundary or specimen reactor environment conditions are created. The historic MARK-III sodium flow loop that achieved thermalhydraulic prototypic test conditions serves as an example of the more complex experiment design package.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Overview of the US Nuclear Science User Facilities

Specialized and often unique and expensive instrumentation, adapted for use on radioactive materials, is essential to conduct research in the area of nuclear energy. Much of this equipment is located in facilities dedicated for such work that have been built and maintained over many decades. The Nuclear Science User Facilities (NSUF) was established in 2007 under the US Department of Energy Office of Nuclear Energy (DOE-NE) in order to bring these capabilities together with those innovative ideas from researchers at universities, national laboratories, small business, and industry that might not have regular access to reactors, hot cells, transportation casks, radiologically qualified instrumentation, and associated expertise. Early descriptions of the originally named Advanced Test Reactor National Scientific User Facility have been given. Through a competitively reviewed proposal process, the NSUF offers access to these capabilities and expertise at no cost to the user to promote research that will address the mission of DOE-NE. As a national organization comprised of a number of partner facilities across the US and located at National Laboratory, university, and industry sites, the NSUF supports materials science studies directed at irradiation effects in nuclear fuels and materials as well as in-pile and post irradiation examination (PIE) instrumentation development. The NSUF has created a Nuclear Fuels and Materials Library (NFML) of previously irradiated materials that users may access for their studies. Expanding the library for future applications of advanced technologies and novel concepts is an important objective of the NSUF. The NSUF has established and maintains the Nuclear Energy Infrastructure Database (NEID) as a tool for both DOE-NE and the nuclear energy research community. The extensive catalogue of capabilities is intended to aid DOE-NE, NSUF, and researchers more effectively and efficiently utilize the available capabilities. (authors)

36 MATERIALS SCIENCE↗

DOE OSTI2018

Reducing Uncertainty in Hydrodynamic Modeling of ATR Experiments Via Flow Testing, Validation, and Optimization

The development, characterization, and qualification testing of nuclear fuel at Idaho National Laboratory’s Advanced Test Reactor (ATR) requires extensive design and analysis activities prior to the insertion of an irradiation experiment in-pile. Significant effort is made in the design and development phase of all in-pile experiments to ensure that the maximum feasible impacts of all necessary experimental requirements are satisfied. The advancement of fuel, cladding, and in-reactor materials technology in recent years has introduced complexities associated with the design and construct of in-pile experiments necessitating deeper understanding of boundary conditions and increasingly comprehensive observations resulting from the experiment. Each unique experiment must be assessed for neutronics response, thermal/hydraulic/hydrodynamic performance, and structural integrity. This is accomplished either analytically, computationally, or experimentally, or some combination thereof, prior to insertion into the ATR. The various effects are interrelated to various degrees, such as the case with the experiment temperature affecting the thermal cross section of the fuel or the increased temperature of the experiment’s materials reducing the mechanical strength of the assemblies. Additionally, the feedback between the experiment’s response to a reactor transient could alter the neutron flux profile of the reactor during the transient. Each experiment must therefore undergo a barrage of analyses to assure the ATR operational safety review committee that the insertion and irradiation of the experiment will not detrimentally affect the safe operational envelope of the reactor. In many cases, the nuclear fuel being tested can be double-encapsulated to ensure safety margins are adequately addressed, whereas failed fuel would be encased in a protective capsule. In other cases, the experiments can be inserted in a self-contained loop that passes through the reactor core, remaining isolated from the primary coolant. In the case of research reactor fuel, however, the fuel plates must be tested in direct contact with the reactor coolant, and being fuel designed for high neutron fluxes, they are inherently power-dense plates. The combination of plate geometry, high-power density, and direct contact with primary coolant creates a scenario where the neutronic/thermomechanic/hydrodynamic characteristics of the fuel plates are tightly coupled, necessitating as complete characterization as possible to support the safety and programmatic assessments, thus enabling a successful experiment. This paper explores the efforts of the U.S. High-Performance Research Reactor program to thermomechanically/hydromechanically characterize the program’s wide variety of experiments, which range from stacks of miniplate capsules to full-sized, geometrically representative curved plates. Special attention is given to instances where the combination of experimental characterization and analytical assessment has reduced uncertainties of the safety margins, allowing experiments to be irradiated that would otherwise not have passed the rigorous qualification process for irradiation in the ATR. In some cases, the combined processes have exposed flow and heat transfer characteristics that would have been missed using historical methods, which allows for more accurate and representative postirradiation assessments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗