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At least 397 records · Page 22
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Hardened Containment Vent System Dose Considerations for FLEX

Fukushima lessons learned include adding diverse and flexible mitigation strategies, or FLEX, that will increase defense-in-depth for beyond-design-basis (BDB) scenarios to address an Extended Loss of AC Power (ELAP). FLEX and NRC order EA-13-109 require improvements to the reliability of Hardened Containment Vent Systems (HCVS) for Mark I and II boiling water reactors (BWR) for venting of the core under severe accident conditions. NEI HCVSWP-02 provides generic guidance to determine radiological dose from HCVS piping. This paper presents a more refined analysis tool and method to assess the doses from the HCVS piping resulting from a failed core during an ELAP for a representative Mark I BWR. The dose rates up to 4' from the HCVS pipe at 8 hours after ELAP and over time at 3' are provided for the different cases. HCVS pipe length can affect dose rates farther from the pipe. As expected, dose rates versus time for representative plant case 2 drop off quickly as the HCVS activity is vented out. The dose rate results show that significant protection for personnel would be required in the areas near HCVS piping during certain time frames of containment venting. Equipment in the vicinity of the vent piping should also be evaluated for impact of radiation exposure. The results for the representative plant cases are higher and lower than the scaled NEI results depending on the model assumptions, limitations and considerations which can change as the HCVS design evolves. Using the NEI paper with scaling factors may not be conservative for all applications.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

A simulation of I 2 S-LWR selected transients

The Integral Inherently Safe (I 2 S-LWR) reactor was designed to improve the safety performance of PWR types reactors, and several design basis accidents were eliminated using innovative approaches to the I 2 S-LWR plant layout. However, some accidents such as station blackout (SBO), main steam line break (MSLB) type accident – the secondary hot leg break (SHLB) – were deemed statistically relevant, and the work performed here was to simulate these events using the system thermal–hydraulics code RELAP5 coupled to the 3D spatial kinetics code PARCS. The safety response of the reactor to a station blackout is predicted to be acceptable at both beginning of cycle (BOC) and end of cycle EOC core conditions with either the UO 2 or U 3 Si 2 fuel options considered here. In the analysis of the SHLB performed here, an unprotected (no SCRAM) condition was assumed to highlight the I 2 S-LWR capabilities. Results indicate that with the UO 2 core, the fuel temperatures rise notably, but no fuel melting was predicted and departure from nucleate boiling ratio (DNBR) limits were not exceeded. Similarly, the U 3 Si 2 fuel performed well, but with lower fuel temperatures and decreased DNBR. Following the break of the steam line in the I 2 S-LWR, shutdown is achieved after about two minutes via negative reactivity from loss of primary flow. In all cases considered, the I 2 S-LWR appears to accomplish the goals of increased safety performance. Further design improvements and other fuel options may marginally improve on the observed safety performance. Additionally, higher fidelity analysis with the next generation subchannel, Computational Fluid Dynamics (CFD), or full core transport may be useful in refining the prediction of the peak fuel and cladding temperatures, as well limiting DNBR during the accidents.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2026

Radiological Releases from Novel Fuel Forms in Advanced Reactors During Severe Accidents for Consequence Analyses

Various advanced reactor developers are exploring the potential for reductions in the size of physical security forces and emergency planning zones. These reductions are based on robust fuel forms and inherently safe reactor designs. However, such reductions in physical protection measures could increase the risk of sabotage. To assess the possibility of reducing these measures, sabotage-induced radiological consequence analyses were carried out. These analyses considered accident scenarios that were beyond design basis accidents and overly conservative (Shah, 2025a; Shah, 2025b; Shah and Hartanto, 2026), yielding very large release fractions. These fractions, which can be used to evaluate physical protection and emergency planning requirements, have been crudely determined and applied as demonstrations for a sodium-cooled fast reactor (SFR) (Shah and Hartanto, 2025a), a high-temperature gas-cooled reactor (HTGR) (Shah and Hartanto, 2025b), a heat pipe–cooled reactor (HPR) (Shah and Hartanto, 2025c), and a molten salt–cooled reactor (MSR) (Shah et al., 2026). A Sandia National Laboratories (SNL) team used MELCOR—a fully integrated severe accident analysis code—to demonstrate the code’s capability to analyze advanced (i.e., not light water–cooled) reactors (including a fluoride salt–cooled high-temperature reactor [FHR]) and calculate radiological releases to the environment during severe accidents (Wagner et al., 2022a, 2022b, 2022c, 2023a, and 2023b). Although the analyses were carried out to demonstrate MELCOR’s growing capability, the release source terms were estimated for advanced reactors, providing valuable insights into the accident progression and radiological releases. These findings from prior SNL studies, including estimated source terms and related sensitivity studies, were leveraged to derive source terms for postulated sabotage-induced accidents. Insights from these sensitivity studies informed the scaling of SNL’s estimated source terms for the defined accident scenarios. The derived release fractions for the severe accident scenarios for the respective reactor designs can be used to perform more nuanced dose consequence analyses to evaluate the reactors’ physical protection and emergency planning zone requirements. These analyses are in accordance with the risk-informed, performance-based approach proposed under 10 CFR Part 53. This study builds on the prior source term analyses and associated sensitivity studies by SNL to derive time-dependent and design-informed release fractions. Section 2 describes the diverse advanced reactor designs analyzed by the SNL team. Section 3 discusses the severe accident analyses, the release fractions calculated, and the limitations and assumptions of the demonstration project. Section 4 presents the release percentages derived for the hypothetical sabotage-induced severe accidents at the advanced reactors. Section 5 summarizes the study’s findings and conclusions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2010

SLUDGE WASHING AND DEMONSTRATION OF THE DWPF FLOWSHEET IN THE SRNL SHIELDED CELLS FOR SLUDGE BATCH 6 QUALIFICATION

Prior to initiating a new sludge batch in the Defense Waste Processing Facility (DWPF), Savannah River National Laboratory (SRNL) is required to simulate this processing, including Chemical Process Cell (CPC) simulation, waste glass fabrication, and chemical durability testing. This report documents this simulation for the next sludge batch, Sludge Batch 6 (SB6). SB6 consists of Tank 12 material that has been transferred to Tank 51 and subjected to Low Temperature Aluminum Dissolution (LTAD), Tank 4 sludge, and H-Canyon Pu solutions. Following LTAD and the Tank 4 addition, Liquid Waste Operations (LWO) provided SRNL a 3 L sample of Tank 51 sludge for SB6 qualification. Pu solution from H Canyon was also received. SB6 qualification included washing the sample per LWO plans/projections (including the addition of Pu from H Canyon), DWPF CPC simulations, waste glass fabrication (vitrification), and waste glass characterization and chemical durability evaluation. The following are significant observations from this demonstration. Sludge settling improved slightly as the sludge was washed. SRNL recommended (and the Tank Farm implemented) one less wash based on evaluations of Tank 40 heel projections and projections of the glass composition following transfer of Tank 51 to Tank 40. Thorium was detected in significant quantities (>0.1 wt % of total solids) in the sludge. In past sludge batches, thorium has been determined by Inductively Coupled Plasma-Mass Spectroscopy (ICP-MS), seen in small quantities, and reported with the radionuclides. As a result of the high thorium, SRNL-AD has added thorium to their suite of Inductively Coupled Plasma-Atomic Emission Spectroscopy (ICP-AES) elements. The acid stoichiometry for the DWPF Sludge Receipt and Adjustment Tank (SRAT) processing of 115%, or 1.3 mol acid per liter of SRAT receipt slurry, was adequate to accomplish some of the goals of SRAT processing: nitrite was destroyed to below 1,000 mg/kg and mercury was removed to below the DWPF target with 750 g of steam per g of mercury. However, rheological properties did not improve and were above the design basis. Hydrogen generation rates did not exceed DWPF limits during the SRAT and Slurry Mix Evaporator (SME) cycles. However, hydrogen generation during the SRAT cycle approached the DWPF limit. The glass fabricated with the Tank 51 SB6 SME product and Frit 418 was acceptable with respect to chemical durability as measured by the Product Consistency Test (PCT). The PCT response was also predictable by the current durability models of the DWPF Product Composition Control System (PCCS). It should be noted, however, that in the first attempt to make glass from the SME product, the contents of the fabrication crucible foamed over. This may be a result of the SME product's REDOX (Reduction/Oxidation - Fe{sup 2+}/{Sigma}Fe) of 0.08 (calculated from SME product analytical results). The following are recommendations drawn from this demonstration. In this demonstration, at the request of DWPF, SRNL caustic boiled the SRAT contents prior to acid addition to remove water (to increase solids concentration). During the nearly five hours of caustic boiling, 700 ppm of antifoam was required to control foaming. SRNL recommends that DWPF not caustic boil/concentrate SRAT receipt prior to acid addition until further studies can be performed to provide a better foaming control strategy or a new antifoam is developed for caustic boiling. Based on this set of runs and a recently completed demonstration with the SB6 Waste Acceptance Product Specifications (WAPS) sample, it is recommended that DWPF not add formic acid at the design addition rate of two gallons per minute for this sludge batch. A longer acid addition time appears to be helpful in allowing slower reaction of formic acid with the sludge and possibly decreases the chance of a foam over during acid addition.

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

Control of Radioactive Methyl Iodide Gases by Mesoporous Hybrid Adsorbents at High Humid Condition - 19329

During the operation of nuclear facility as well as decommissioning stage, various radioactive gaseous wastes are generated. It is generally recognized that radioiodine is potentially one of the most hazardous gas on the human body and environment. Iodine-131 release in the gaseous effluents at the nuclear facilities had been a critical environmental concern. Reducing iodine release is a critical issue in radioactive waste system evaluation, especially in their adherence to the 'as low as reasonably achievable' (ALARA) criteria because radioiodine among these gaseous wastes has radiological harm. Therefore, numerous studies for the removal of radioiodine from the off-gas stream of nuclear facilities have been performed with formation processes and chemical structure. There has been some concern regarding the use of TEDA-impregnated activated carbon as an adsorbent for radioiodine removal, because of the low ignition point of TEDA and the possible release of TEDA from the charcoal in continuously flowing air streams. Therefore, the TEDA-impregnated activated carbon cannot use to trap radioiodine generated from high humid condition, due to the rapid drop of the removal efficiency. It is obvious that high humid can be expected during design basis event such as loss-of-coolant accident (LOCA). Therefore, investigation on the retention of organic radioiodine (CH3I131) by various activated carbons at high humid condition has extensively been performed so far. But there are no good results on the adsorption of methyl iodide at high humid condition up to now. As a result, the promising approach to the problem of iodine removal under high humid conditions is to use organic-inorganic hybrid adsorbents in which are enough retention characteristics from hydrophobic property. In order to develop the best removal efficiency for the control of radioiodine gas generated from a high humid process, a hydrophobic adsorbents was developed and the performance test of adsorbent using radioactive methyl iodide tracer was also performed. Demonstration test of adsorbent using radioactive methyl iodide tracer was performed at high humid conditions up to 97% relative humidity. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Scaling for Nuclear Reactor System: Overview

A scaling analysis is essential for predicting the behavior of a reactor system under various operating conditions, as well as for ensuring safe and reliable experimentation and adequate dataset generation for supporting the system code model development and assessment in accordance with USNRC guidelines 1.203. Testing and modeling can be carried out at various scales, from small test units to full-scale reactors, to validate the predictions produced by the analysis. However, due to the cost and challenges associated with full-scale testing, the nuclear industry uses scaled test facilities—integral effect test (IET) and separate effect test (SET)—to identify potential design issues and optimize the design to ensure regulatory compliance for safer operation and anticipated accident management. In general, IET facilities are non-nuclear thermal hydraulics test facilities. In contrast, SET facilities can be non-nuclear testing (e.g., steam generator, containment, pump system SETs) and nuclear testing (mostly fuel-material irradiation testing). IETs are used to study the system-level thermal-hydraulic behavior of the reactor coolant system (RCS) under different anticipated operational occurrences (AOOs) and design basis accidents (DBAs) such as loss-of-coolant-accident (LOCA) and loss-of-power (i.e., station blackout) and loss-of-flow (i.e., failure of RCS pump). Several LOCA scenarios could happen in reactor DBAs, such as main steam line break (MSLB), SG tube rapture (SGTR), break in direct vessel injection (DVI), and other small breaks LOCA. Scaling analysis determines the appropriate size and operating conditions for the test facility. This involves analyzing the system's thermal-hydraulic behavior at different scales and identifying the key parameters that affect the thermal-hydraulic behavior of the reactor. These key parameters can be identified so that the scaled facility can replicate the expected conditions in a prototypical setting. The scaled facility can simulate the same flow rates, pressure gradients, heat transfer coefficients, and fluid properties as the prototype. Scaling analysis is used to ensure that the results from these tests can be scaled up to predict the behavior of a full-scale prototype with an acceptable range of scaling distortions.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

DWPF Simulant CPC Studies For SB8

Prior to processing a Sludge Batch (SB) in the Defense Waste Processing Facility (DWPF), flowsheet studies using simulants are performed. Typically, the flowsheet studies are conducted based on projected composition(s). The results from the flowsheet testing are used to 1) guide decisions during sludge batch preparation, 2) serve as a preliminary evaluation of potential processing issues, and 3) provide a basis to support the Shielded Cells qualification runs performed at the Savannah River National Laboratory (SRNL). SB8 was initially projected to be a combination of the Tank 40 heel (Sludge Batch 7b), Tank 13, Tank 12, and the Tank 51 heel. In order to accelerate preparation of SB8, the decision was made to delay the oxalate-rich material from Tank 12 to a future sludge batch. SB8 simulant studies without Tank 12 were reported in a separate report.1 The data presented in this report will be useful when processing future sludge batches containing Tank 12. The wash endpoint target for SB8 was set at a significantly higher sodium concentration to allow acceptable glass compositions at the targeted waste loading. Four non-coupled tests were conducted using simulant representing Tank 40 at 110-146% of the Koopman Minimum Acid requirement. Hydrogen was generated during high acid stoichiometry (146% acid) SRAT testing up to 31% of the DWPF hydrogen limit. SME hydrogen generation reached 48% of of the DWPF limit for the high acid run. Two non-coupled tests were conducted using simulant representing Tank 51 at 110-146% of the Koopman Minimum Acid requirement. Hydrogen was generated during high acid stoichiometry SRAT testing up to 16% of the DWPF limit. SME hydrogen generation reached 49% of the DWPF limit for hydrogen in the SME for the high acid run. Simulant processing was successful using previously established antifoam addition strategy. Foaming during formic acid addition was not observed in any of the runs. Nitrite was destroyed in all runs and no N2O was detected during SME processing. Mercury behavior was consistent with that seen in previous SRAT runs. Mercury was stripped below the DWPF limit on 0.8 wt% for all runs. Rheology yield stress fell within or below the design basis of 1-5 Pa. The low acid Tank 40 run (106% acid stoichiometry) had the highest yield stress at 3.78 Pa.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2011

IMPACT OF SMALL COLUMN ION EXCHANGE STREAMS ON DWPF GLASS FORMULATION MELT RATE STUDIES

This study was undertaken to evaluate the potential impacts of the Small Column Ion Exchange (SCIX) streams - particularly the addition of Monosodium Titanate (MST) and Crystalline Silicotitanate (CST) - on the melt rate of simulated feed for the Defense Waste Processing Facility (DWPF). Additional MST was added to account for contributions from the Salt Waste Processing Facility (SWPF). The Savannah River National Laboratory (SRNL) Melt Rate Furnace (MRF) was used to evaluate four melter feed compositions: two with simulated SCIX and SWPF material and two without. The Slurry-fed Melt Rate Furnace (SMRF) was then used to compare two different feeds: one with and one without bounding concentrations of simulated SCIX and SWPF material. Analyses of the melter feed materials confirmed that they met their targeted compositions. Four feeds were tested in triplicate in the MRF. The linear melt rates were determined by using X-ray computed tomography to measure the height of the glass formed along the bottom of the beakers. The addition of the SCIX and SWPF material reduced the average measured melt rate by about 10% in MRF testing, although there was significant scatter in the data. Two feeds were tested in the SMRF. It was noted that the ground CST alone (ground CST with liquid in a bucket) was extremely difficult to resuspend during preparation of the feed with material from SCIX and SWPF. This feed was also more difficult to pump than the material without MST and CST due to settling occurring in the melter feed line, although the yield stress of both feeds was high relative to the DWPF design basis. Steady state feeding conditions were maintained for about five hours for each feed. There was a reduction in the feed and pour rates of approximately 15% when CST and MST were added to the feed, although there was significant scatter in the data. Analysis of samples collected from the SMRF pour stream showed that the composition of the glass changed as expected when MST and CST were added to the feed. These reductions in melt rate are consistent with previous studies that showed a negative impact of increased TiO{sub 2} concentrations on the rate of melting. The impact of agitating the melt pool via bubbling was not studied as part of this work, but may be of interest for further testing. It is recommended that additional melt rate testing be performed should a potential reduction in melt rate of 10-15% be considered an issue of concern, or should the anticipated composition of the glass with the addition of material from salt waste processing be modified significantly from the current projections, either due to changes in sludge batch preparation or changes in the composition or volume of SCIX and SWPF material.

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

Waste Chemistry Envelope for the Determination of Pitting Corrosion Susceptibility - 19554

Liquid radioactive wastes at the Hanford and Savannah River sites are stored in large underground carbon steel tanks. The high-level wastes are processed in several of the tanks and then transferred by piping to other site facilities for further processing before they are stabilized in a vitrified or grout waste form. Based on waste removal and processing schedules, many of the tanks, will be required to be in service for times exceeding the initial intended life (i.e., greater than 60 years). Until the waste is removed from storage, transferred, and processed, the materials and structures of the tanks must maintain a confinement function by providing a barrier to the environment and by maintaining acceptable structural stability during design basis events, which include loadings from both normal service and abnormal (e.g., earthquake) conditions. Administrative programs are in place to mitigate potential corrosion mechanisms and thereby maintain the structural and leak integrity functions of these waste tanks throughout their intended service life. Waste chemistry and temperature control are the means by which corrosion degradation of the waste tanks is minimized. Chemistry controls are based on in-situ coupon tests, waste tank experience, and laboratory testing. These requirements comprise the technical standards for corrosion control. The standards are utilized then as guidelines for operational procedures. The present waste chemistry controls for the Hanford DSTs were established in the 1980's in response to tank failures caused by stress corrosion cracking (SCC) at the DOE Savannah River Site (SRS). Portions of the SRS chemistry limits, which were developed at the Savannah River National Laboratory (SRNL) were coupled with corrosion testing results from Pacific Northwest National Laboratories (PNNL) and adopted for corrosion control of the double-shell tank (DST) waste at the Hanford Site. Since that time, the temperatures of the waste in the DSTs have decreased to less than 50 deg. C in most cases as the radioactive material has decayed. As a consequence, new chemistry controls were formulated for SCC in 2010. This testing however, did not address that potential for pitting in the supernatant and at the liquid-air interface. Indeed, initial tests indicated that the inhibitor requirements for SCC would be insufficient for the prevention of pitting. In addition to lower waste temperatures, future retrieval and process changes may impact the DST waste chemistry. Initial projections suggest that the waste chemistry of the DSTs may shift to a broader range of pH and higher aggressive anion concentrations (e.g., chloride, sulfate, etc.) than the wastes that are presently stored. Additionally, processes within the vitrification facility may result in streams that are returned to the tank farm for evaporation. These streams may also contain high concentrations of aggressive species. SRNL investigated the propensity for pitting of the carbon steel waste tanks at current and projected DST waste chemistries. A statistically designed series of tests provided a pitting factor that relates the ratio of the inhibitor species to the aggressive species and the probability of observing pitting. An acceptable pitting factor was determined and was utilized for the development of a new chemistry envelope. The new chemistry envelope was compared to the envelope that was established for SCC and was determined to be bounding. Validation of the new chemistry envelope was performed via model simulations and comparison with historical experimental results and approaches to establishing chemistry envelopes. The new chemistry envelope will eventually be part of an overall assessment of the condition of the tanks and any changes made to the overall structural integrity program for the Hanford DSTs. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Design of a supervisory control system for autonomous operation of advanced reactors

Advanced reactors to be deployed in the coming decades will face deregulated energy markets, and may adopt flexible operation to boost profitability. To aid in the transition from baseload to flexible operation paradigm, autonomous operation is sought. This work focuses on the control aspect of autonomous operation. Specifically, a hierarchical control system is designed to support constraint enforcement during routine operational transients. Within the system, data-driven modeling, physics-based state observation, and classical control algorithms are integrated to provide an adaptable and robust solution. A 320 MW Fluoride-cooled High-temperature Pebble-bed Reactor is the design basis for demonstrating the proposed control system. The hierarchical control system consists of a supervisory layer and low-level layer. The supervisory layer receives requests to change the system's operating conditions (e.g., the current reactor power to meet a load -follow), and accepts or rejects them based on constraints that have been assigned. Constraints are issued to keep the plant within an optimal operating region. The low-level layer interfaces with the actuators of the system to fulfill requested changes, while maintaining tracking and regulation duties. Further, to accept requests at the supervisory layer, the Reference Governor algorithm was adopted. To model the dynamics of the reactor, a system identification algorithm, Dynamic Mode Decomposition, was utilized. To estimate the evolution of process variables that cannot be directly measured (e.g., the propagation of delayed neutron precursors), the Unscented Kalman Filter, incorporating a nonlinear model of nuclear dynamics, was adopted. The composition of these algorithms led to a numerical demonstration of constraint enforcement during a 40% power drop transient (at a rate of 5 %/min). Uncontrolled secondary-side temperatures were successfully constrained. Adaptability of the proposed system was demonstrated by modifying the constraint values, and enforcing them during the transient. Robustness was also demonstrated by enforcing constraints under noisy environments.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2010

Tc-99 Adsorption on Selected Activated Carbons - Batch Testing Results

CH2M HILL Plateau Remediation Company (CHPRC) is currently developing a 200-West Area groundwater pump-and-treat system as the remedial action selected under the Comprehensive Environmental Response, Compensation, and Liability Act Record of Decision for Operable Unit (OU) 200-ZP-1. This report documents the results of treatability tests Pacific Northwest National Laboratory researchers conducted to quantify the ability of selected activated carbon products (or carbons) to adsorb technetium-99 (Tc-99) from 200-West Area groundwater. The Tc-99 adsorption performance of seven activated carbons (J177601 Calgon Fitrasorb 400, J177606 Siemens AC1230AWC, J177609 Carbon Resources CR-1240-AW, J177611 General Carbon GC20X50, J177612 Norit GAC830, J177613 Norit GAC830, and J177617 Nucon LW1230) were evaluated using water from well 299-W19-36. Four of the best performing carbons (J177606 Siemens AC1230AWC, J177609 Carbon Resources CR-1240-AW, J177611 General Carbon GC20X50, and J177613 Norit GAC830) were selected for batch isotherm testing. The batch isotherm tests on four of the selected carbons indicated that under lower nitrate concentration conditions (382 mg/L), Kd values ranged from 6,000 to 20,000 mL/g. In comparison. Under higher nitrate (750 mg/L) conditions, there was a measureable decrease in Tc-99 adsorption with Kd values ranging from 3,000 to 7,000 mL/g. The adsorption data fit both the Langmuir and the Freundlich equations. Supplemental tests were conducted using the two carbons that demonstrated the highest adsorption capacity to resolve the issue of the best fit isotherm. These tests indicated that Langmuir isotherms provided the best fit for Tc-99 adsorption under low nitrate concentration conditions. At the design basis concentration of Tc 0.865 µg/L(14,700 pCi/L), the predicted Kd values from using Langmuir isotherm constants were 5,980 mL/g and 6,870 mL/g for for the two carbons. These Kd values did not meet the target Kd value of 9,000 mL/g. Tests conducted to ascertain the effects of changing pH showed that at pH values of 6.5 and 7.5, no significant differences existed in Tc-adsorption performance for three of the carbons, but the fourth carbon performed better at pH 7.5. When the pH was increased to 8.5, a slight decline in performance was observed for all carbons. Tests conducted to ascertain the temperature effect on Tc-99 adsorption indicated that at 21 ºC, 27 ºC, and 32 ºC there were no significant differences in Tc-99 adsorption for three of the carbons. The fourth carbon showed a noticeable decline in Tc-99 adsorption performance with increasing temperature. The presence of volatile organic compounds (VOCs) in the source water did not significantly affect Tc-99 adsorption on either of two carbons tested. Technetium-99 adsorption differed by less than 15% with or without VOCs present in the test water, indicating that Tc-99 adsorption would not be significantly affected if VOCs were removed from the water prior to contact with carbon.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2023

Development of a BlueCRAB/MELCOR Framework for Supporting Realistic Mechanistic Source Term Calculations in Microreactors

Efforts are currently underway to deploy microreactor modeling and simulation tools to better support vendors and regulatory authorities in submitting and reviewing licensing applications. In particular, the Nuclear Regulatory Commission is expected to rely on the Comprehensive Reactor Analysis Bundle (BlueCRAB) multiphysics toolset in performing design- and beyond-design-basis accident analyses. In addition, the Nuclear Regulatory Commission has been using the MELCOR code to estimate mechanistic source terms during accidents. As MELCOR relies on isotopic inventory and reactor temperature/power evolution profiles during accident conditions—all of which can theoretically be obtained from BlueCRAB—the ultimate goal of this activity is to establish a common BlueCRAB-MELCOR framework. However, prior to the present research, BlueCRAB had never been used to calculate such quantities of interest at the full-core level. While there are many Monte Carlo (MC) codes capable of computing such quantities of interest, they are unable to readily account for multiphysics feedback. BlueCRAB allows for the coupling of different physics codes together to perform multiphysics-informed calculations. Therefore, the purpose of this fiscal year 2023 work is to investigate the feasibility and challenges of performing such calculations within BlueCRAB so as to generate the data that MELCOR relies on. To demonstrate the methodology, the proposed workflow was applied to a prototypical heat pipe-cooled microreactor model. To predict isotopic concentrations (taking into account the ac- cumulation of fission products during operation), the necessary microscopic cross sections were generated via OpenMC and tabulated with respect to temperature and burnup. Next, a recently developed capability in Griffin (the reactor physics application in BlueCRAB) was used to convert the OpenMC output format into the ISOXML format used by Griffin. A multiphysics microscopic depletion calculation that involved performing a coupled full-core, heterogeneous neutron trans- port and thermal calculation at each depletion step was conducted to deplete the core to end of life (EOL) conditions so as to provide both isotopics and the initial condition for the transient calculation. Following a brief null-transient to verify that the initial condition had been properly restarted and was indeed in thermal equilibrium, a heat pipe failure transient was simulated. Thus, the entire workflow of using BlueCRAB to generate MELCOR inputs, from cross-section generation to producing isotopic inventory and power/temperature evolution profiles during transients, is demonstrated. This report also details the identified gaps in the workflow and how they were (for the most part) addressed. Future work should focus on directly including MEL- COR into the workflow by performing a MELCOR calculation using the BlueCRAB-generated input data. In addition, the heat pipe reactor design should be improved so as to reflect more prototypical burnup characteristics at EOL.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Rethinking of Zircaloy Embrittlement Criteria for Improving Safety Margin of Light Water Reactor

Zircaloy embrittlement criteria have become a basis for the current emergency core cooling system (ECCS) criteria to maintain the coolable geometry. U.S. NRC has considered that retaining adequate cladding ductility is a simple and effective way for ensuring the structural integrity of fuel rods under the accident conditions. Considering this, the current ECCS criteria limits the peak cladding temperature(PCT) to be 1204 deg. C, and equivalent cladding reacted (ECR) to be 17%, calculated by the Baker-Just equation. However, the criteria were established based on the experimental investigations in early 1970's and it has inevitably limitations, in particular, about hydrogen embrittlement with burnup. As a consequence, the new draft-rule proposes a relation that gives rise to a decreasing ECR limit with respect to pre-transient hydrogen content by U.S NRC. The regulation of Japan is quite different from that of U.S. NRC. This rule is set to assure structural integrity of ruptured fuel rods upon water-quenching. Recently, Japanese Atomic Energy Agency (JAEA) conducted integral fuel rod thermal shock tests with high burnup claddings. These investigations confirmed that the original criteria (PCT 1200 deg. C, ECR 15% calculated by Baker-Just equation ) still successfully guarantee fuel rod structural integrity upon thermal shock and concluded that the high-burnup effects are inconsiderable as the thermal-shock fracture boundary is not significantly reduced. Hence, there exists an obvious disagreement between the embrittlement-based and the thermal shock fracture-based criterion particularly when it comes to the high burnup consideration. The one of them can be adopted to define safety limit and safety margin which is the difference between the regulation limit and actual value of the parameter. Also, it is worth noting that the temperature limit (1204 deg. C) of the Zircaloy embrittlement criteria has not been altered since it was first determined in 1970's. Such a qualitative concern alone should not limit further discussions on the potential increase in the temperature limit. Indeed, there is ample experimental evidence that supports the possibility that Zircaloy ductility can be maintained beyond 1204 deg. C if it is oxidized for a limited time period. The increase in the temperature margin can lead to a reactor power uprate, or allow an increased accident coping time. This paper aims to achieve a comprehensive understanding of an engineering compatibility between nuclear fuel behaviors, safety system, and safety criteria, in the pursuit of increasing safety margin of LWRs. Specifically, this study presents a revision the current temperature limit (1204 deg. C) with adjustable ECR levels through an extensive literature review. And the impact of this temperature limit on the actual design basis accident margin is discussed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

FEED Study of CarbonCapture Inc DAC and CarbonCure Utilization Technologies Using United States Steel’s Gary Works Plant Waste Heat (Final Report)

The University of Illinois at Urbana-Champaign (UIUC) led this project to produce a front-end engineering design (FEED) study of an advanced Direct Air Capture and Utilization System (DACUS) system that can remove a minimum of 5,000 tonnes/yr net of carbon dioxide from air (based on cradle-to-gate LCA) and utilizing the CO 2 to produce low carbon intensity ready mix concrete. The designed system, if built, would be larger than any currently existing Direct Air Capture (DAC) collector in the U.S. Such carbon capture technologies are critical to meeting the goals of the DOE’s program to accelerate climate-critical technology. In addition to the power sector, industrial facilities for the manufacture of steel and cement/concrete are among the major sources of anthropogenic CO 2 . DAC is a promising new technology for reducing CO 2 , a potent greenhouse gas, in the atmosphere but is expensive, in part due to the energy required to adsorb and desorb captured CO 2 during cycles. By integrating CarbonCapture Inc. (CCI) DAC modules at United States Steel's Gary Works (USS) and utilizing the site's waste heat, energy, and location this project evaluates the feasibility of utilizing the captured CO 2 and the logistics of transportation. CarbonCapture Inc. has developed an innovative DAC system using novel adsorbents to cost-effectively capture CO 2 . The captured, liquified gas will be trucked to ready-mix concrete plants within the region, the closest of which is approximately 3.5 miles away, where CarbonCure will inject it into concrete during the mixing process at the facilities. The carbon dioxide reacts with concrete, mineralizing into calcium carbonate (CaCO 3 ), permanently locking the greenhouse gas into the matrix of the building material. This FEED study demonstrated a full CO 2 value chain for DACUS from industrial facilities. It also provided a means for Visage Energy Corp. (Visage) to assess the impact of this holistic approach on job creation, regional economic impact, and environmental justice. The project team also included Sargent & Lundy (S&L) to provide the constructability review and costing of the integration of the DAC with the steel plant. Ecotek Engineering USA, LLC designed the outside battery limit (OSBL) infrastructure to connect the DAC and the plant. Activities performed during the project included: (1) Project Management Plan; (2) Technology Maturation Plan (TMP); (3) Initial Workforce Readiness Plan; (4) Workforce Readiness Plan; (5) Front-End Engineering Design (FEED) Study; (6) Project Design Basis; (7) Hazards and Operability (HAZOP) Study; (8) Constructability Review; (9) Project Cost Assessment; (10) Logistics Analysis of CO 2 Transportation to the Utilization Site; (11) Business Case Analysis; (12) Life Cycle Analysis (LCA); (13) Environmental Health and Safety (EH&S) Analysis; (14) Environmental Justice Analysis; and (15) Economic Revitalization and Job Creation Outcomes Analysis. This report provides a summary of the outcomes and results of the project, which was performed between Oct. 1, 2022, through Sept. 30, 2024.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2017

Operator Action Time Evaluation of Emergency Water Injection Strategy for Extreme SGTR Accidents

In terms of severe accidents caused by earthquakes or tsunamis that are beyond expectation, a special safety inspection for operating plants, following the Fukushima accident, has been conducted by the government of Korea to verify that nation's nuclear power plants have been adequately designed to respond to extreme accidents. The inspection results show that Korean nuclear power plants in operation have no imminent risk for an expected maximum potential earthquake or coastal flooding based on the up-to-date investigation. However, there is a need to implement long- and short-term safety improvements in order to secure the level of safety for natural beyond-design-basis events. One of the measures to increase the mitigation capability during extreme accidents is installing the injection flow paths to provide emergency cooling water of extemal sources using fire engines to the steam generators (SGs) or reactor cooling system (RCS). An emergency cooling water system consists of a fixed pipe connected from the RCS or SGs to the outside of the containment. A standby valve is installed on the pipe. Following the occurrence of an extreme accident, movable equipment (for example, a fire truck hose) can be connected to the pipe hole at the opening of the isolation valve. The emergency cooling water injection strategy was not taken into account for the steam generator tube rupture (SGTR) accident management strategy of the OPR-1000. Therefore, it is necessary to develop some guidelines or strategies to cope with an extreme SGTR scenarios using the newly installed injection flow paths and fire engines. As long as the fission products released from the core during severe accidents are kept inside the containment, there may be several ways to reduce the radionuclide release into the environment, for example, by using plant safety features. The containment bypass sequences, however, allow the fission products to be released into the environment or directly to the auxiliary building. The SGTR accidents, which are one of the containment bypass sequences, have shown during various probabilistic safety assessments (PSAs) to be contributors to plant risk, mainly because of their potential for causing a release outside containment. This paper is intended to be an investigation of the characteristics of accident progressions and assessments for the mitigative effectiveness of the external cooling water injection strategies during extreme SGTR accidents in a typical OPR-1000 plant. Analyses were performed using a MAAP computer code version 5.03. The MAAP code is a system-level computer code capable of conducting integral analyses of potential severe accident progressions in nuclear power plants, whose main purpose is to support a level 2 PSA or severe accident management strategy developments.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Leveraging operational information from wastewater injection wells to evaluate CO 2 injection performance for carbon storage applications in the Appalachian Basin

Abstract Geologic parameters, geophysical logging, injection testing, and operational metrics from wastewater injection wells were integrated to develop a preliminary design of a carbon storage facility in the Appalachian Basin. A scattered group of 10–20 commercial wastewater injection wells dispose off produced water from oil and gas wells in the region, utilizing a sequence of stacked deep saline formations for injection zones. These wastewater injection wells provide practical benchmarks for understanding the feasibility of carbon dioxide (CO 2 ) storage. Geologic models were developed based on characterization data from the wastewater injection wells. Reservoir simulations were calibrated according to injection testing and operational data from the wastewater injection wells. Long‐term operational data on injection flow rates and pressures measured in the wastewater injection wells were especially useful to evaluate the performance of carbon storage applications. The simulations were used to estimate injection pressures, radius of CO 2 saturation, and pressure response for industrial scale CO 2 storage applications. Results were also used to provide a design basis in terms of number of injection wells, well spacing, area of review, injection system components, monitoring plan, and CO 2 pipeline distribution system. The analysis demonstrates that there is sufficient injectivity in the deep saline formations in the west‐central Appalachian Basin to store commercial volumes of anthropogenic CO 2 . The geologic system appears suitable for supporting CO 2 injection rates of 0.5–1.0 million metric tons per year at injection pressures below formation fracture pressure in a single well. The long‐term operational data of wastewater injection wells within the study area suggested a lower permeability‐thickness values than indicated by initial reservoir tests. A workflow for developing realistic permeability values for input into reservoir simulations is presented. © 2020 Society of Chemical Industry and John Wiley & Sons, Ltd.

Valluri, Manoj↗
DOE OSTI2018

Structure of the Public Irradiation Dose During Operation of Experimental-Demonstration Power Complex Enterprises

The structure of the public radiation dose resulting from gas aerosol emissions of the experimental-demonstration energy complex incorporating the BREST-OD-300 reactor and fuel fabrication and reprocessing modules is determined. The yearly radiation dose to humans at the point of maximum ground concentration of radionuclides from design basis emissions of the energy complex is formed primarily by {sup 3}H,{sup 14}C, and fission products (0.73, 1.2 2, and 0.9 μSv, respectively) through the peroral pathway as a result of the fallout during the running year of operation of the enterprises. The largest contribution to the irradiation dose is due to the emissions from the spent fuel reprocessing module – 2.33 μSv/yr – and 50% is due to {sup 14}C. The radiation dose from the emissions of the BREST-OD-300 reactor is almost completely due to {sup 3}H and {sup 210}Po (0.73 and 0.17 μSv/yr, respectively). The emissions from the fuel fabrication module have the smallest effect on the public. The dose from fission products is produced by emissions from the reprocessing module approximately in equal amounts in terms of external and internal pathways (0.47 and 0.43 μSv).

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗