Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “SHUTDOWN”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 1,351 records · Page 75
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Flowsheet Definition for the Tank Side Cesium Removal Unit and the Importance of Early Communication - 19142

The U.S. Department of Energy (DOE) tank operations contractor Washington River Protection Solutions, LLC (WRPS) is working toward the start of low-activity waste (LAW) treatment in direct feed mode as soon as 2022. WRPS is responsible for providing the pre-treated feed for this direct feed mode and had originally planned to use the Low Activity Waste Pretreatment System (LAWPS) to supply the initial feed in 2021. In September 2017, in response to a DOE-sponsored LAWPS program review, a new system was proposed to provide the initial feed during direct feed mode until a supplemental pretreatment capability is ready to begin processing feed. This system, known as the Tank Side Cesium Removal (TSCR) system, is based on the Savannah River Tank Closure Cesium Removal (TCCR) system and technology used to separate cesium out of reactor water at Fukushima. A preliminary flowsheet was needed to form the initial basis for future decisions and allow discussions between operations, project management, and the customer. The main goal of the TSCR flowsheet was to form a process flow diagram, process description, and much of the preliminary information to be used to help inform future decisions. When flowsheet development first began, little to no information was readily known and what was known needed to be compiled in one place and digestible by several audiences. The impacts of various parameters on the system also needed a vehicle for clear communication to varying audiences. While a more complicated model was being completed, a model of the process was created using Microsoft Excela which could have several parameters changed while maintaining basic logical requirements (such as treated waste volume cannot exceed the volume of the Interim LAW Storage Tank [ILST], must shutdown for so many days for column changeout, etc.). This program allowed the unit downtime, start dates, treatment rates, and several other variables to be modified and reflect the ILST storage volume as a function of mission progression. The model with several different parameters gathered from the various groups was combined into an early flowsheet of the TSCR process and created one of the first citable documents about TSCR implementation at Hanford. (authors)

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

Molten Salt Reactor Experiment (MSRE) Lay-Up - 19187

UCOR, an AECOM-led partnership with Jacobs, manages the cleanup of the 2,200-acre East Tennessee Technology Park for its client, the U.S. Department of Energy (DOE), in Oak Ridge, Tennessee. The site was contaminated with radioactive, hazardous and industrial wastes generated by more than 40 years of national defense and energy missions. 2019 is the 50. anniversary of the shutdown of the Molten Salt Reactor Experiment (MSRE). MSRE ran a brief four years in the 1960's but earned an enduring legacy as an innovative reactor technology concept which has enjoyed a renewal of global interest in the past several years. In 2018, it was given elevated status among advanced reactor programs within DoE's Nuclear Energy office. The MSRE was a graphite-moderated, liquid fueled reactor that operated from June 1965 through December 1969. Although it was shut down nearly 50 years ago, certain systems within the reactor have continued to operate to maintain the facility in a safe and stable condition until it can be demolished as part of the comprehensive cleanup of the Oak Ridge Reservation. These systems are composed of a complex network of pipes, valves, gauges and associated equipment that control air filtration and pressure buildup resulting from fluorine gas generated through radiolysis of the fluoride salt which once fueled the reactor. In 2016, an MSRE Engineering Evaluation Team was chartered to assess current facility conditions and develop recommendations to maintain the facility in a safe state until deactivation and demolition commences. Plans are now underway to 'lay up' the MSRE facility until deactivation and demolition can occur. The proposed MSRE Facility Layup will minimize maintenance costs, reduce personnel risk of injury and exposure, provide reliable electric service to the facility and require zero personnel occupancy at MSRE. In the proposed layup approach a new source of electrical power to required loads will be routed through the facility, and new gas processing and ventilation systems will be installed that will be optimized for safe processing of the constituent hazardous gasses that exist in the legacy vessels at MSRE. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Temporal Analysis and Scene Change Detection in Multispectral Overhead Imagery

Scene change detection can be a tedious and time consuming process especially when concerning large geographical areas, and the process can be even more cumbersome when analyzing changes in an area over large spans of time. Developing a useful way to help analysts recognize at what points in time significant changes to a scene have occurred can allow them to better focus their efforts in characterizing events. Applications include: Facility monitoring, Construction chronology, Monitoring of vehicle/aircraft activity, Characterization of larger sequences of events. In large areas exceeding hundreds to thousands of square kilometers in size, it can be difficult localizing when scene changes have occurred. Analysts can spend hours going through imagery to try to identify new construction, monitor facility activities, monitor vehicle movement, etc. where the object of interest may only be a few square meters. Our goal is to help cut down this time by giving analysts change maps with hot spots of change, allowing them to focus on regions that have experienced actual change in time frames they're interested in. Additionally, by combining these change maps into layers within a data cube, analysts can examine the change maps from a temporal perspective, allowing events to be characterized over spans of time. By opening the data cube in an imaging software capable of separating the layers, we can analyze the change maps sequentially, allowing us to examine scene changes occurring over time. As an example, we examined overhead imagery from Planet Labs of what appears to be a parking lot on Fort Irwin over the course of a year using ENVI, a geospatial satellite imagery analysis software. Using ENVI, we generate a graph of changes over time, and notice a particular segment near the end of our analysis window where no changes are detected. Examination of the actual satellite imagery reveals that during this time span, the parking lot was empty. This could be due to facility shutdown for maintenance or upgrades, or possibly even total workforce/vehicle fleet movement. Information like this could help analysts better characterize events, as well as to help create clearer timelines in larger sequences of events. Workflow steps: - Collect multiple maps of the same AOI (Area of Interest) during a time span of interest; - Generate change maps from AOI maps; - Generate data cube from change maps. An analyst can use the data cube to help inspect an AOI for activities within a time span of interest. If an event of interest is discovered, the analyst can then refer to the maps corresponding to the appropriate dates and times in the data cube to see precisely what is transpiring. The biggest objective being worked on is improving the change detection methodology employed. We currently use PCA-EM (Principal Component Analysis with Expectation Maximization), but we are currently focusing on implementing IR-MAD (Iteratively Reweighted Multivariate Alteration Detection) to be used in conjunction with PCA-EM in an effort to decrease false positivity and noise in the change maps we generate.

42 ENGINEERING↗
DOE OSTI2020

Filter 'Cook-Off': Filter Selection via Side-by-Side Testing at Actual Process Conditions - 20040

In 2005, the THOR steam reforming fluidized-bed reactor technology was selected by the U.S. Department of Energy for treatment of sodium-bearing waste at the INL. During the final phase of commissioning prior to radiological operations, regeneration of the high temperature filters in the process gas filter (PGF) vessel became ineffective. After approximately 50 days of testing, the differential pressure across the filters began increasing exponentially and the plant had to be shutdown. Subsequent investigation revealed that the Inconel 625 (I625) sintered metal filter media was corroding in the 550 deg. C reducing environment. A pilot plant run tested four candidate replacement filter media. The top candidate was further tested during a long-term pilot plant run to validate the selection. Data from the testing indicate the selected filters will provide reliable service throughout the life of the project, including after in-situ radiological decontamination. This paper presents the data from this effort and reinforces the value of pilot plant testing. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Reducing Cost of Chlorinated Volatile Organic Compound Remediation by Transitioning from Active to Passive Soil Vapor Extraction - 20157

Areas of high chlorinated volatile organic compound (cVOC) contamination at the Savannah River Site (SRS) have been undergoing remediation via soil vapor extraction, sometimes coupled with thermal treatments to enhance extraction rates. These active systems are effective in removing large amounts of contaminant mass from the subsurface and mitigating the impacts to groundwater. However, as extraction rates decline, costs must be evaluated with respect to the benefit of continued active operation. A decision framework for identifying conditions when a transition to a more passive remediation is appropriate has been developed with state and federal regulatory agencies. Two remediation areas have recently been transitioned from active soil vapor extraction (ASVE) to passive soil vapor extraction (PSVE) at the SRS. Performance evaluation goals including plume stabilization, overall mass removal trends, environmental sustainability and costs were considered in transitioning from active remediation to passive technologies at both sites. At the Dynamic Underground Stripping (DUS) project at the M-Area Settling Basin, ASVE was combined with steam injection to extract cVOCs during active operations. Steam injection occurred from September 2005 to September 2009. DUS utilized 63 steam injection wells, 34 active vapor extraction wells, and 3 active soil vapor extraction units (SVEUs). Two active SVEUs had 60 horsepower blowers; the third had a 25-horsepower blower. Mass removal was closely tracked during DUS operations; over 181,437 kilograms (400,000 pounds) of cVOCs were removed while active steam injection occurred. After steaming was stopped, ASVE continued. The 34 active wells were evaluated in 2012. The ASVE wells were grouped into categories of high, medium, and low extraction rates. High producing wells remained connected to a single active SVEU. Low producing wells were abandoned, and the medium producing wells were transitioned to PSVE (Microblowers{sup TM}). Microblowers{sup TM} utilize a dedicated blower per well and are solar powered. This passive technology provides energy, maintenance, and operation costs savings while still providing an efficient reduction in cVOC migration to groundwater. In 2018, the remaining ASVE wells were evaluated again. The purpose of this testing was to identify which wells removed the most mass. An optimal well configuration was determined. The criteria to discontinue ASVE was removal of less than 18 kilograms (40 pounds) per week of cVOCs. After 3 months of shutdown (rebound conditions), 2.7 kilograms (5.9 pounds) of cVOCs per week were being removed. Data from the rebound test justified ending ASVE and transitioning wells with higher extraction rates to PSVE. Wells that had depleted the cVOC mass within their zone of influence were abandoned. Performance data from existing PSVE wells justified ending PSVE at wells with depleted extraction rates. Currently the system has 16 PSVE wells operating. Another ASVE system was being used to treat cVOC contaminated soil at the A-Area Miscellaneous Rubble Pile (AMRP) at SRS. System operation began in 2004, with 7 ASVE wells connected to a 60- horsepower blower. Mass removal rates and contaminant concentrations remained consistently low over the ASVE lifespan at AMRP. This indicated that mass removal was diffusion limited. With this data, the 7 ASVE wells were transitioned to PSVE in 2017. Twelve pressure monitoring points were also transitioned to PSVE wells. AMRP currently has 19 PSVE wells operating. Both transitions from active to passive remediation had concurrence from the United States Environmental Protection Agency and the South Carolina Department of Health and Environmental Control. These transitions ensure that only the necessary amount of energy is being exerted to remediate the environment. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

TCCR Operational Summary and Optimization for Tank 9 Processing - 20314

Savannah River Remediation (SRR) manages and operates the liquid waste facilities at Savannah River Site (SRS) for the Department of Energy (DOE). Stored liquid waste is a complex mixture of insoluble solids (sludge) and soluble salts in an alkaline solution. SRR has deployed the Tank Closure Cesium Removal (TCCR) system, a tank-side ion exchange process, to remove radioactive cesium from salt waste and enable onsite disposal of the resulting decontaminated salt solution as low-level waste at the Saltstone facilities. The TCCR system consists of two prefilters, four ion exchange (IX) columns, one resin trap, and a ventilation system. The IX process uses a form of inorganic crystalline silicotitanate (CST), which has a high affinity for cesium and other alkali metals, strontium, and actinides. This process is currently deployed utilizing salt feed from Tank 10, with future plans to dissolve solid salt in Tank 9 and transfer the salt solution to Tank 10 for processing through TCCR. The feed for TCCR must be created from salt-cake in Tank 10 through a dissolution process. Once enough salt has been dissolved, a qualification process is entered. This process characterizes the feed and ensures the cesium loading on the columns will not cause boiling of waste within the columns during or after processing. Once the batch has been qualified, salt waste is fed to the TCCR system through a transfer pump in the center of the tank. The waste is filtered through a set of two shielded, dead-end prefilters that prevent solids buildup in the columns. The filtered salt solution then travels to the shielded IX columns, which can be operated individually or in series, where the cesium is sorbed on the CST media. The decontaminated salt solution (DSS) then travels through a resin trap and out of the module to Tank 11. TCCR has successfully processed approximately 795,000 L of Tank 10H radioactive salt waste over two batches to date. There has not yet been a system induced shutdown. The prefilters performed as expected with only minor degradation in recovery of differential pressure after a backflush sequence. The time between backflushes decreased as each batch reached the end of processing. The hydraulics in the IXCs mostly performed as expected at all flow rates, except for one IXC that will be further investigated during Batch 3 processing. The TCCR system has shown some opportunities for more efficient processing during the length of the demonstration so far. For future processing of material from Tank 9H through Tank 10H and the TCCR unit, TCCR 1A will implement changes to the prefilters and the IXCs. The prefilters will have an increased surface area and a new filter media in an effort to increase time between filter swaps and improve backwashing cleaning capability. The IXCs will have a reduced diameter to allow for increased heat transfer out of the column and increased loading of Cs-137. Additionally, a new form of CST with an increased kinetic performance is being investigated for use during TCCR 1A operation. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Influence of protective actions on ingress-water mass in steam generator tube rupture of HTGR - 18479

Steam generator tube rupture(SGTR) accident is one characteristic accident of High Temperature Gas-Cooled Reactor (HTGR), which may cause a potential hazard to reactor system. Because the amount of water ingress into reactor core is related closely to the severity of the accident, the research on water-ingress is significant for the evaluation on inherent safety characteristics of HTGR. Both different accident scenario and the system design will have influence on the amount of the ingress-water under SGTR. In this paper, a pebble-bed high temperature gas cooled reactor is modeled by thermal-hydraulic system analysis code and the water-ingress transient under SGTR is simulated to study the influence of the protective action of emergency shutdown on the water ingress. The research indicates that the protective action has no obvious influence on the amount of the ingress-water within the current studied range. The amount of ingress-water is controlled below the design value to satisfy the demand of reactor safety. The analysis results will provide some support for the HTGR research. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
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 OSTI2016

In-Line Method for Calculating Peak Neutron Absorption by Fission Products Following a Change in Reactor Power

When the power level in an operating reactor is reduced the concentration of xenon-135 and samarium-149 temporarily increases as the nuclides (and their immediate precursors) equilibrate to the new power level. During this period, the increased inventory of these fission products causes an insertion of negative reactivity that can last for tens of hours. Understanding and predicting this transient reactivity effect is important in reactor design and operation. Explicit depletion calculations are often used to predict the reactivity behavior following a power change. However, these calculations require many time-steps to fully resolve the reactivity trend, and are correspondingly expensive. In 2010 an in-line method for estimating the peak xenon mass in a core following shutdown was described. However, there is no guarantee that the peak negative reactivity condition for a core will occur at peak xenon mass. Therefore, in this summary, we extend this original work using a new method that can predict the peak neutron absorption rate in the core due to {sup 135}Xe and {sup 149}Sm following any reduction in power. The proposed method can also be used to calculate the peak positive reactivity transient due to an increase in reactor power, but this scenario is of less interest in most reactor design applications. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Experimental Study of Molten Metallic Fuel Relocation in Sodium-Filled Fuel Structures

The Prototype Generation-IV Sodium-cooled Fast Reactor (PGSFR) is being developed by Korea Atomic Energy Research Institute (KAERI). While their design incorporates inherent safety features demonstrated in the landmark Shutdown Heat Removal Test (SHRT) conducted at EBR-II including a large pool configuration and uranium alloy fuel, it is still necessary to evaluate metal fuel relocation and theirs consequences in case of postulated core disruptive accidents. Extensive data regarding fuel relocation in sodium from both in-pile and out-of-pile tests were obtained for oxide fuels in the 1960's - 1970's, while few studies of metallic fuel behavior were performed because of concerns that metallic fuel could not achieve high burnup due to the irradiation induced swelling problem. Following new metal fuel design development that mitigated the swelling problem, experiments regarding interaction of molten metal fuels and liquid sodium have been conducted by several researchers. Most studies that were conducted using metal simulants in various liquids concluded that the molten metals were heavily fragmented after liquid contact and the dominant factors governing the fragmentation were superheating and latent heat of fusion of the molten metal fuel. Gabor et al. conducted experiments using kilogram quantities of various molten uranium alloys in an open sodium pool configuration. They found that the metal fuel fragments were basically in forms of filaments and sheets with a high bed voidage in the order of 0.9. Their calculation indicated that the debris beds formed by relocated metal fuels in an open sodium pool would be largely coolable by conduction heat transfer; and even if deep beds were to form, convective heat transfer and boiling heat transfer could preclude further melt penetration. Although Gabor's tests are useful in evaluating the coolability of relocated metal fuels during a postulated accident in a SFR, they focused on late phase behavior in which the core material relocates downward through the core support plate into the inlet plenum where the structures are more open and thus consistent with the open pool configuration. There is still a knowledge gap related to nature of metal fuel fragmentation and relocation within the fuel assembly structures. Argonne National Laboratory (ANL) is planning to conduct experiments regarding metal fuel relocation in the core structures which mock up the PGSFR's fuel assembly design. In the present study, preliminary tests for demonstration of melting and injecting the metal fuel were conducted using copper. Metal fuel relocation experiments with single fuel pin test section are planned to be conducted for understanding the flow and freezing behavior of the molten metal fuel in a narrow sodium channel and for evaluating the coolability of the relocated fuel. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Used Nuclear Fuel Management in Asia

Used fuel discharged from nuclear reactors are presently stored in wet pools or in dry cask at-reactor and/or away-from- reactor sites. Only a few countries reprocess their civilian used fuel, notably France, Russia, and UK. No country has yet operated a geologic repository to dispose of civilian used fuel or high-level radioactive wastes (HLW). As a result, the management and disposal of used fuel and HLW are viewed as the Achilles' heel of nuclear power. Figure 1 shows the used fuel inventory in the United States (US) and East Asian nuclear power programs. The inventory is estimated based on that in 2010 and the projected growth of nuclear power in these programs to 2020 and 2030. Figure 1 indicates that in 2010, the US had about 63,900 tons of used fuel from light-water reactors (LWR), more than the statutory limit of what was supposed to go into the now-abandoned Yucca Mountain Repository, and accounting for more than 33% of global used LWR fuel total. Currently, about 2,000 tons of used fuel are generated in the existing US LWRs annually. Russia had 25,100 tons of used fuel in 2010. Based on a modest growth of nuclear power, its used fuel inventory is expected to increase to about 31,800 and 38,000 tons by 2020 and 2030, respectively. The nuclear energy programs in East Asia (China, Japan, ROK, and Taiwan) account for 19% of global used LWR fuel inventory in 2010. China's used fuel inventory in 2010 was relatively small at 1,900 tons, but is expected to grow rapidly to {approx}12,400 and {approx}31,000 tons in 2020 and 2030, respectively, based on its expanding nuclear power program (from the current installed capacity of 21.8 GW to {approx}60 and {approx}200 GW in 2020 and 2030, respectively). Besides PWRs, China also operates 2 Candu reactors which each produces roughly 8 times more used fuel in tons than a PWR of the same power. The used Candu fuel inventory would increase from 950 tons in 2010 to 2,850 and 4,750 tons in 2020 and 2030, respectively. Used fuel in Japan, ROK, and Taiwan are US-obligated. Japan's future trends of used fuel inventory are uncertain due to the unknown numbers and timing of the currently-shutdown nuclear power plants that would be allowed for restart, and the uncertain future of Japan's reprocessing at Rokkasho. Under two scenario cases (a no-nuke/no-reprocessing and a 80%-nuke with reprocessing), the Japanese used fuel inventory could be capped from 2020 to 2030 to {approx}21,500 tons for the first case, and to {approx}22,500 tons for the second. The ROK's used PWR fuel inventory would increase from 4,900 tons in 2010 to about 8,800 and 14,000 tons in 2020 and 2030, respectively. Its used Candu fuel inventory would also increase from 5,900 tons in 2010 to about 9,700 and 13,500 tons in 2020 and 2030, respectively. The Taiwan used fuel inventory should be kept at current level as nuclear power is being phase out. Its growth should be modest from 3,000 tons in 2010 to about 4,300 and 5,000 tons in 2020 and 2030, respectively even if the 4. nuclear power station would start and the operation of existing LWRs is extended. Used fuel inventory in other East Asian countries (DPRK, Vietnam, etc.) should be small, if any before 2030.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Preliminary Analysis of FFTF Loss-of-Flow Tests with SAS4A/SASSYS-1

The Fast Flux Test Facility (FFTF) was the most recent Sodium-cooled Fast Reactor (SFR) built in the United States. The Passive Safety Test (PST) program was carried out on FFTF between 1984 and 1986 to demonstrate the capability of passive reactor shutdown and decay heat removal in response to protected and unprotected transients. In order to recover the information and knowledge that has been gained from FFTF, a research project was initiated focusing on the simulation of Unprotected Loss of Flow Without Scram (LOFWOS) tests similar to the knowledge preservation activities conducted for EBR-II by Argonne National Laboratory in earlier years. These knowledge preservation activities are intended to support the validation of computational tools and methods in the area of SFR passive safety. Comparisons with experimental data and other safety codes provide opportunities to improve SFR design and safety analysis codes and methods.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Effect of Delayed Energy Release on Power Normalization in Reactor Depletion Calculations

Reactor depletion calculations typically assume that the thermal power of the reactor is directly proportional to the instantaneous fission rate within the core. However, approximately 7% of the sensible energy released per fission comes from the radioactive decay of fission products, which occurs alter the fission event. Thus, the assumed direct relationship between reactor power and fission rate implies that all fission products have reached secular equilibrium within the core. In situations where fission product concentrations have not reached equilibrium (e.g., shortly after reactor startup or a change in reactor power during operation), the delayed energy release due to fission product decay creates a nonlinear relationship between reactor power and fission rate. This non-linear relationship, in turn, affects the neutron flux levels and fuel consumption rate required to maintain the desired thermal power output for the reactor. For example, during initial startup of a commercial light water reactor the apparent energy release per fission is ∼7{sup %} lower than the typical value of ∼200 MeV/fission due to the potential energy stored in radioactive fission products. As a consequence, the flux level (and, therefore, uranium consumption rate) in the reactor must be correspondingly higher to preserve the desired reactor power. As the reactor operates and fission product inventories increase, the decay heat produced by the reactor will rise, producing a corresponding increase in the apparent energy release per fission. The characteristic time-dependent behavior of energy release due to the decay of fission products (i.e., decay heat) following reactor startup, shutdown, or changes in reactor power is well-known and widely-used in reactor safety analyses. However, to our knowledge, the effects of time-dependent decay heating rates on reactor depletion calculations have not been thoroughly investigated. In this paper, we present a simple methodology for explicitly accounting for delayed energy release due to the decay of fission products in reactor depletion calculations. In the proposed method, the balance equations for fission product decay energy are incorporated into the depletion coefficient matrix, using a strategy similar to that described by Isotalo in 2016. The modified system of differential equations can be solved via a matrix exponential approximation (or other suitable numerical method) to yield solutions for both nuclide densities and decay energy emission densities as a function of time. Numerical results from the method are presented for simplified (one-group point (0-D) and 1-D) representations of a commercial pressurized water reactor (PWR), in order to estimate the effect of delayed energy release on fuel consumption and flux level immediately following reactor startup.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

OECD/NEA EGMPEBV Activities in Multi-Physics Verification and Validation

The Expert Group on Multi-Physics Experimental Data, Benchmarks and Validation (EGMPEBV) was established in 2014 by the Nuclear Energy Agency (NEA) at the Organization for Economic Co-operation and Development (OECD) to establish processes for certification of experimental data and development of benchmark models for validation of Multi-Physics (MP) Modeling and Simulation (M and S) computational systems. The aim of the group is to provide member countries with guidelines and recommendations for validating and improving their novel multi-physics simulations, and with an access to key experimental data. To be used to their full potential, these novel tools will require a more complex suite of validation experiments due to the multiple energy, spatial and temporal scales, as well as the number of physical phenomena being simulated. They will also require accurate measurements of all terms, including coupling terms representing the feedback effects, in situations where the appropriate experimental techniques and facilities might not exist today. However, the ability to conduct validation experiments has for certain applications progressed at a slower pace than for computational methods, or in some cases it has significantly regressed (through shutdown of facilities, or through retirement of experts). In addition, as modelling capabilities are reaching deeper levels in single physics phenomena and in coupled behaviors, appropriate experimental techniques may not yet exist. Validation of multiple physics models requires a wide variety of experimental data, which emphasizes the importance of maximizing the use of historically accumulated data to avoid the significant cost of performing similar experiments today. Moreover, the pedigree and quality of this data as it pertains to modern day simulation methods may be insufficient for rigorous validation of high fidelity modeling and simulation tools. Thus, these data can often exhibit limitations including a lack of measurement accuracy or a lack of knowledge of the measurement uncertainty in the experimental techniques; an absence of measurements for some critical parameters; an obscuring of relationships between individual physics phenomena owing to the output data resolution; or simply an inability to allow for a detailed reinterpretation of the experiments based on the documented information. The organization of the EGMPEBV relies on three Task Forces, focused on 1) Experimental Data Qualification and Benchmark Evaluation; 2) Methods and Standards, and 3) Specific Applications. Task Force 1 is focused on providing better and more accurate experimental datasets, which support validation of high-fidelity multi-physics modelling and simulation (M and S) tools. At the same time it has to consider the incompleteness of past experimental data sets (in terms of data, documentation, or uncertainties) and with the limited number of available multi-physics experimental facilities. Task Force 2 is designed to develop validation methods and guidelines, and uncertainty qualification for the new family of multi-physics, multi-scale codes in the context of emerging demands such as longer fuel cycles and power uprate. Task Force 3 is focused on applications of validation experiments and has the objective to organize multi-physics M and S benchmarks involving validation experiments.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

Investigation of Radiation Cross-Talk Effects on Biological Dose Rates in Maintenance Areas Inside the ITER Tokamak - Paper 107

The intense plasma neutron source of ITER requires careful consideration of radiation safety during the design of the machine, its operation plan and licensing process. One important issue is that of occupational radiation exposure (ORE) during in-situ maintenance activities inside the cryostat, and particularly in the port inter-space area (PI) where up keeping of port systems should take place (diagnostics, tritium breeding test modules, heating antennas). ITER ports and port plugs (PP) become activated by neutron irradiation during plasma operation; from the subsequent decay and gamma emission of irradiated materials, the so-called shutdown dose rate field (SDR) appears. A project requirement of 100 uSv/h after 12 days exists for the SDR in the PI in order to limit exposure during maintenance and meet ORE regulatory limits. This requirement is particularly challenging for equatorial ports (EP) due to, among others, background neutron radiation and activation from/by neighbouring components (a.k.a. 'cross-talk'), and particularly the lower ports (LP). For a particular port inter-space, the SDR field is a combination of: (i) the gamma dose due to the activation of components local to the PI, typically referring to the PI itself plus the port plug, port duct and bellows, vessel extension and blankets in its vicinity; (ii) the gamma dose due to activation of components far away from the PI, a.k.a. the cross-talk, which typically refers to other ports, cryostat and rest of the vacuum vessel and blankets. The situation is complicated to some extent by the fact that both these components can arise from a combination of irradiation by local neutrons, i.e. those permeating through the vicinity of the port itself, and by cross-talk neutrons, i.e. those permeating via other paths. In simple terms, the so-called 'neutron cross-talk' refers to neutron leakage through other ports and components producing activation, decay gammas and dose locally in the PI of interest, whereas the 'gamma cross-talk' refers to activation and decay gammas in other ports producing dose in the PI from far away. The study reported here focuses on the evaluation of the effect of radiation cross-talk from a torus cryopump (TCP) LP located directly underneath a generic diagnostics equatorial port plug (DEPP) in the equatorial port itself. This is considered the worst case scenario, as the TCP-LP has a large aperture and has to be mostly empty in order to meet TCP pumping requirements. For the reference case and R1, the neutron cross-talk from the LP activating the EP is responsible for 109 uSv/h. In addition, a gamma cross-talk of 51 uSv/h from LP activation has been found. The total cross-talk contribution is therefore responsible for 160 uSv/h of SDR at EP PI. The local SDR contribution, i.e. that from activation of the EP with 'local-only' neutrons, is 166 uSv/h. The total SDR at R1 is 360 uSv/h, including some minor contributions from the UP, lateral EPs, cryostat and bio-shield. The total SDR at R2 is 349 uSv/hr, with a cross-talk contribution of 203 uSv/h. In R2 the gamma cross-talk is more intense, as it is closer to the highly activated TCPH. These results illustrate how the SDR field produced by radiation cross-talk through a TCP-LP is comparable in magnitude with that produced by neutron leakage and activation local to the EP. Shielding options are currently under consideration. Local gamma shielding at the EP PI is found to be effective in reducing gamma cross-talk from the LP and TCPH, and also in reducing the local contribution from the port duct and bellows. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2015

Comsol multiphysics simulation of the 1958 Y-12 criticality accident - 14654

This paper presents the simulation of the 1958 Y-12 criticality excursion using the commercial multiphysics code COMSOL Multiphysics. ICNC 2011 reported the initial effort to simulate the Y-12 accident power history using a fully coupled neutronic and computational fluid dynamic (CFD) approach with the Imperial College London FETCH code. That simulation included the initial spike and a subsequent power history 30 seconds afterwards. A time-dependent lumped parameter excursion model was later developed at Imperial in November 2014 and taken to 20 min. The COMSOL simulation is viewed as a hybrid treatment of these efforts. The use of COMSOL as a platform for criticality excursion simulation had been previously demonstrated with SILENE step reactivity insertion benchmarks. The Y-12 accident represented a unique time dependent reactivity insertion mechanism by dilution of the fissile concentration over time that facilitated the excursion termination. The COMSOL model includes a moving mesh vertical boundary, compartment radiolytic gas model, coupled turbulent fluid flow and conjugate heat transfer. The results are compared with both FETCH and lumped parameter simulations. The COMSOL power history results are in agreement with the reported and reconstructed initial spike fissions, power and total fissions over time. The simulated power trace shows explicit subsequent spikes approximately 10-20 seconds apart from the initial spike, over the first three minutes of the excursion. This result qualitatively confirms the radiation dose rate trace from a far field monitor located well beyond the affected facility. Additionally, the COMSOL model shows the onset of boiling after 3 minutes, as postulated in the literature. The combination of boiling and fissile solution dilution contributed to the shutdown mechanism. (authors)

42 ENGINEERING↗
DOE OSTI2022

Activation of lead coolant in Westinghouse lead fast reactor

The Westinghouse Lead-cooled Fast Reactor (LFR) is a medium-sized, passively safe, scalable reactor being developed by Westinghouse in collaboration with domestic and international organizations with the primary goals of reducing front-end capital cost and generating flexible and cost-competitive electricity while satisfying the highest standards in terms of safety and sustainability. An important aspect of new plant design is an assessment of activation level of its components so that an appropriate nuclear waste management plan can be established. As lead coolant is a distinguishing design feature of the LFR, its activation due to irradiation and the implications on establishing an appropriate waste disposal approach are important for engagement of project stakeholders, customers and nuclear safety regulators. This paper provides a brief description of the Westinghouse LFR design, a background on lead activation phenomenology, and then describes the methodology used for lead activation analysis performed with the computer code Serpent 2. It was determined that impurities naturally occurring in lead have the biggest contribution to lead activation. Corrosion and erosion products from the components immersed in lead can also be important contributors to lead activity and their impact can be limited by appropriate selection of materials and operating conditions. Polonium isotopes, resulting from bismuth generated from lead irradiation and its impurities, bring a notable contribution to the inhalation and ingestion doses during the plant operation and in the first years after the plant shutdown. However, it is not by far as important as for lead-bismuth coolant.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

INL Soil Contamination Areas - Wildland Fire Radiological Hazards

The largest wildland fire on the INL occurred in 2019 and initiated a reassessment of the hazard of wildfires burning through soil contamination areas. In 2020 during the COVID shutdown, the INL Emergency Management Group and the Radiological Control Group worked together to re-evaluate the hazards from these soil contamination areas that were last evaluated in 2001. The new soil sample data was examined, and the areas were mapped for radiation intensity. A new evaluation of the radiological hazards due to wildfire was completed and issued because of this work. This presentation will describe the work and the methodologies used to complete this re-evaluation.

61 RADIATION PROTECTION AND DOSIMETRY↗