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DOE OSTI2020

Performing a multi-unit level-3 PSA with MACCS

MACCS (MELCOR Accident Consequence Code System), WinMACCS, and MelMACCS now facilitate a multi-unit consequence analysis. MACCS evaluates the consequences of an atmospheric release of radioactive gases and aerosols into the atmosphere and is most commonly used to perform probabilistic safety assessments (PSAs) and related consequence analyses for nuclear power plants (NPPs). WinMACCS is a user-friendly preprocessor for MACCS. MelMACCS extracts source-term information from a MELCOR plot file. The current development can combine an arbitrary number of source terms, representing simultaneous releases from a multi-unit facility, into a single consequence analysis. The development supports different release signatures, fission product inventories, and accident initiation times for each unit. The treatment is completely general except that the model is currently limited to collocated units. A major practical consideration for performing a multi-unit PSA is that a comprehensive treatment for more than two units may involve an intractable number of combinations of source terms. This paper proposes and evaluates an approach for reducing the number of calculations to be tractable, even for sites with eight or ten units. The approximation error introduced by the approach is acceptable and is considerably less than other errors and uncertainties inherent in a Level 3 PSA.

42 ENGINEERING↗

DOE OSTI2020

An Overview of the State-of-the-Art Reactor Consequence Uncertainty Assessment Accident Progression Insights

The U.S. Nuclear Regulatory Commission (NRC) with Sandia National Laboratories (Sandia) have completed three uncertainty analyses (UAs) as part of the State-of-the-Art Reactor Consequence Analyses (SOARCA) program. The SOARCA UAs included an integrated evaluation of uncertainty in accident progression, radiological release, and offsite health consequence projections. The UA for Peach Bottom, a boiling-water reactor (BWR) with a Mark I containment located in the State of Pennsylvania, analyzed the unmitigated long-term station blackout SOARCA scenario. The UA for Sequoyah, a 4-loop Westinghouse pressurized-water reactor (PWR) located in the State of Tennessee, analyzed the unmitigated short-term station blackout SOARCA scenario, with a focus on issues unique to the ice condenser containment and the potential for early containment failure due to hydrogen deflagration. The UA for Surry, a 3-loop Westinghouse PWR with a sub-atmospheric large dry containment located in the State of Virginia, analyzed the unmitigated short-term station blackout SOARCA scenario including the potential for thermally-induced steam-generator tube rupture. These three UAs are currently documented in three NUREG/CR reports. This report provides input to planned NRC documentation on the insights and findings from the SOARCA UA program. The purpose of the summary report is to provide a useful reference for regulatory applications that require the evaluation of offsite consequence risk from beyond design basis event severe accidents. This report focuses on the accident progression and source term insights developed from the MELCOR analyses. MELCOR is the NRC's best-estimate, severe accident computer code used in the SOARCA UAs. In anticipation of the SOARCA UA insights work, NRC and Sandia benchmarked the response of the Peach Bottom model to selected reference calculations from the Peach Bottom SOARCA UA. Peach Bottom was the first SOARCA UA performed and was completed in 2015 using the MELCOR 1.8.6 code. The PWR SOARCA UAs evolved the original methodology and utilized the updated MELCOR 2.2 computer code. The Peach Bottom model has been systematically updated for other NRC research efforts and has been updated to MELCOR 2.2. computer code. The findings from the new reference calculations using the updated model with the MELCOR 2.2 code are also integrated into the report. A second objective is an assessment of the applicability of the results to the other nuclear reactors in the U.S. As the key findings are reviewed, judgments are presented on the applicability of the results to other U.S. nuclear power plants. An important objective of the SOARCA program relied on high- fidelity plant-specific modeling. However, the nature of the insights and conclusions allowed judgements to be made on the applicability of the various insights to the same general classification of plant (i.e., BWR or PWR) or the entire fleet of plants. Finally, the results from the SOARCA UA accident progression calculations contain a wealth of information not previously documented in the NUREG/CRs. This report includes new but related information that can be used to benchmark past or support future regulatory decisions related to severe accidents. The new work includes a benchmark of the NUREG-1465 licensing source term definitions, the variability of key accident progression events and timing to radionuclide release, and an improved understanding of the timing and source terms from consequential steam generator tube ruptures. iii ACKNOWLEDGEMENTS The Sandia authors gratefully acknowledge the significant technical and programmatic contributions from the NRC SOARCA team which are reflected throughout the report. Dr. Tina Ghosh has been involved throughout the SOARCA UAs, providing the primary managerial and technical oversight. The long lists of NRC and Sandia contributors from the SOARCA UAs are cited in the three NUREG/CRs and are also gratefully acknowledged by the small team of authors compiling the results of their efforts. Significant technical contributions, advice, and reviews were provided by Dr. Hossein Esmaili, Dr. Alfred Hathaway, and Dr. Edward Fuller (retired) of the NRC. Dr. Randal Gauntt (retired), Mr. Patrick Mattie, Mr. Joseph Jones (retired), and Dr. Doug Osborn from Sandia are recognized as the SOARCA UA managers guiding the past efforts. There is a comparable list of project managers at the NRC including Ms. Patricia Santiago, Dr. Salman Haq, and Mr. Jon Barr. Sadly, we have lost Mr. Charlie Tinkler and Mr. Robert Prato, who were important contributors to the original SOARCA project. Finally, Mr. Kyle Ross and Mr. Mark Leonard have also retired but were significant technical contributors. Mr. Kyle Ross was the technical lead on all three SOARCA UAs and the original pressurized water reactor SOARCA study. Mr. Leonard was the technical lead on the original boiling water reactor SOARCA study and a key contributor to the first Peach Bottom SOARCA UA. iv

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

Severe Accident Research Directions after Fukushima

The accidents at the Fukushima Daiichi nuclear power station in March of 2011 renewed interest globally in the nature, causes, and health and economic consequences of severe accidents involving core damage and containment impairment. As a result of these accidents, research has been conducted into the modeling and phenomenological understanding of the accidents at Fukushima and the response and effectiveness of mitigating measures such as emergency water injection, primary system depressurization and containment venting actions. An important example of this is the jointly conducted OECD-NEA organized Benchmark of the Fukushima Accidents (BSAF project) where accident reconstruction analyses using computer codes such as MELCOR and MAAP are ongoing. Even prior to the Fukushima accidents, the Sandia/NRC sponsored State of Art Reactor Consequence Analyses (SOARCA) foreshadowed the accidents in Japan where long-term and short-term station blackout accidents in boiling water reactors with Mark-1 containment were analyzed showing striking similarities to the accidents at Fukushima. These analyses also highlighted important differences in our prior understanding, in particular the importance of suppression pool thermal stratification and the extraordinary operation of the Fukushima Unit-2 reactor core isolation cooling (RCIC) system, prompting addition research into both of these areas. Additionally, the focus of the world's major computer codes on modeling the Fukushima accidents in the BSAF effort, led to several 'Crosswalk' comparisons between code predictions, highlighting subtle but important modeling differences between the codes and identifying areas where modeling improvements can be made. Collectively, these research areas are aimed at improving our understanding of severe accident progression and real-world response of safety related systems and components, and should lead to improved accident response and mitigation capabilities in the future. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2025

Investigating the Role of Accident Tolerant Cladding on Source Term Reduction for High-Burnup PWRs Using MELCOR

The use of accident tolerant fuel (ATF) cladding can increase coping times during and beyond design basis accidents. While such gains may be incremental, they provide a margin that can potentially be recovered to enable high-burnup (HBU) operation. Realizing such a margin requires demonstrating that the combination of HBU and ATF has not led to an overall increase in source term. This study investigates the influence of cladding technology (Zr-based, Cr-coated Zr, and FeCrAl) and fuel cycle length (18 and 24 months) on radiological dose at the boundary of the exclusion zone for a four-loop pressurized water reactor to investigate whether ATF claddings can provide such benefits. We analyze a recovered large break loss-of-coolant accident scenario to investigate the impact of transient timescale on the benefits of such coping time increases. The simulations have been performed using the MELCOR and MELCOR Accident Consequence Code System codes. For the cases analyzed, increased fuel cycle length did not necessarily increase radionuclide release and hydrogen generation, as these were found to be sensitive to the core power distribution. Similarly, off-site dose consequence is dominated by short-lived radionuclides that tend to saturate earlier in the burnup, so higher burnup operation did not necessarily increase the source term for the phenomena and transients analyzed here. Delays in recovery of the lowpressure safety injection system increase hydrogen production and radionuclide release, especially between 780 s and 1620 s, due to the nonlinear oxidation and core degradation behavior. Results show that Cr-coated Zr enhances safety by delaying heatup and gap release. Here, when uncertainty propagation on oxidation properties is considered, FeCrAl exhibits the lowest overall radionuclide release and off-site dose throughout the spectrum. However, while the considered “base model” performance is superior under delayed injection scenarios, upper-bound cases display hydrogen generation risk comparable to the Zr-based cladding.

Accident Tolerant Fuel↗

DOE OSTI2016

A methodology for accident analysis of fusion breeder blankets and its application to helium-cooled lead–lithium blanket

'Fusion for Energy' (F4E) is designing, developing, and implementing the European Helium-Cooled Lead-Lithium (HCLL) and Helium-Cooled Pebble-Bed (HCPB) Test Blanket Systems (TBSs) for ITER (Nuclear Facility INB-174). Safety demonstration is an essential element for the integration of these TBSs into ITER and accident analysis is one of its critical components. A systematic approach to accident analysis has been developed under the F4E contract on TBS safety analyses. F4E technical requirements, together with Amec Foster Wheeler and INL efforts, have resulted in a comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena while remaining consistent with the approach already applied to ITER accident analyses. Furthermore, the methodology phases are illustrated in the paper by its application to the EU HCLL TBS using both MELCOR and RELAP5 codes.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2011

OECD MCCI project long-term 2-D molten core concrete interaction test design report, Rev. 0. September 30, 2002.

The Melt Attack and Coolability Experiments (MACE) program at Argonne National Laboratory addressed the issue of the ability of water to cool and thermally stabilize a molten core-concrete interaction when the reactants are flooded from above. These tests provided data regarding the nature of corium interactions with concrete, the heat transfer rates from the melt to the overlying water pool, and the role of noncondensable gases in the mixing processes that contribute to melt quenching. As a follow-on program to MACE, The Melt Coolability and Concrete Interaction Experiments (MCCI) project is conducting reactor material experiments and associated analysis to achieve the following two technical objectives: (1) resolve the ex-vessel debris coolability issue through a program that focuses on providing both confirmatory evidence and test data for the coolability mechanisms identified in MACE integral effects tests, and (2) address remaining uncertainties related to long-term two-dimensional molten core-concrete interactions under both wet and dry cavity conditions. Achievement of these two objectives will demonstrate the efficacy of severe accident management guidelines for existing plants, and provide the technical basis for better containment designs for future plants. In terms of the first program objective, the Small-Scale Water Ingression and Crust Strength (SSWICS) test series has been initiated to provide fundamental information on the ability of water to ingress into cracks and fissures that form in the debris during quench, thereby augmenting the otherwise conduction-limited heat transfer process. A test plan for Melt Eruption Separate Effects Tests (MESET) has also been developed to provide information on the extent of crust growth and melt eruptions as a function of gas sparging rate under well-controlled experiment conditions. In terms of the second program objective, the project Management Board (MB) has approved startup activities required to carry out experiments to address remaining uncertainties related to long-term two-dimensional molten core-concrete interaction. In particular, for both wet and dry cavity conditions, there is uncertainty insofar as evaluating the lateral vs. axial power split during a core-concrete interaction due to a lack of experiment data. As a result, there are differences in the 2-D cavity erosion predicted by codes such as MELCOR, WECHSL, and COSACO. The first step towards generating this data is to produce a test plan for review by the Project Review Group (PRG). The purpose of this document is to provide this plan.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2011

Final Review of Safety Assessment Issues at Savannah River Site, August 2011

At the request of Savannah River Nuclear Solutions (SRNS) management, a review team composed of experts in atmospheric transport modeling for environmental radiation dose assessment convened at the Savannah River Site (SRS) on August 29-30, 2011. Though the meeting was prompted initially by suspected issues related to the treatment of surface roughness inherent in the SRS meteorological dataset and its treatment in the MELCOR Accident Consequence Code System Version 2 (MACCS2), various topical areas were discussed that are relevant to performing safety assessments at SRS; this final report addresses these topical areas.

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2012

A Recommended Pasquill-Gifford Stability Classification Method for Safety Basis Atmospheric Dispersion Modeling at SRS

Several of the most common methods for estimating Pasquill-Gifford (PG) stability (turbulence) class were evaluated for use in modeling the radiological consequences of SRS accidental releases using the MELCOR Accident Consequence Code System, Ver. 2 (MACCS2). Evaluation criteria included: (1) the ability of the method to represent diffusion characteristics above a predominantly forested landscape at SRS, (2) suitability of the method to provide data consistent with the formulation of the MACCS2 model, and (3) the availability of onsite meteorological data to support implementation of the method The evaluation resulted in a recommendation that PG stability classification for regulatory applications at SRS should be based on measurements of the standard deviation of the vertical component of wind direction fluctuations, σ e , collected from the 61-m level of the SRS meteorological towers, and processed in full accordance with EPA-454/R-99-005 (EPA, 2000). This approach provides a direct measurement that is fundamental to diffusion and captures explicitly the turbulence generated by both mechanical and buoyant forces over the characteristic surface (forested) of SRS. Furthermore, due to the potentially significant enhancement of horizontal fluctuations in wind direction from the occurrence of meander at night, the use of σ e will ensure a reasonably conservative estimate of PG stability class for use in dispersion models that base diffusion calculations on a single value of PG stability class. Furthermore, meteorological data bases used as input for MACCS2 calculations should contain hourly data for five consecutive annual periods from the most recent 10 years.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2012

Recommended Tritium Oxide Deposition Velocity for Use in Savannah River Site Safety Analyses

The Defense Nuclear Facilities Safety Board (DNFSB) has recently questioned the appropriate value for tritium deposition velocity used in the MELCOR Accident Consequence Code System Ver. 2 (Chanin and Young 1998) code when estimating bounding dose (95th percentile) for safety analysis (DNFSB 2011). The purpose of this paper is to provide appropriate, defensible values of the tritium deposition velocity for use in Savannah River Site (SRS) safety analyses. To accomplish this, consideration must be given to the re-emission of tritium after deposition. Approximately 85% of the surface area of the SRS is forested. The majority of the forests are pine plantations, 68%. The remaining forest area is 6% mixed pine and hardwood and 26% swamp hardwood. Most of the path from potential release points to the site boundary is through forested land. A search of published studies indicate daylight, tritiated water (HTO) vapor deposition velocities in forest vegetation can range from 0.07 to 2.8 cm/s. Analysis of the results of studies done on an SRS pine plantation and climatological data from the SRS meteorological network indicate that the average deposition velocity during daylight periods is around 0.42 cm/s. The minimum deposition velocity was determined to be about 0.1 cm/s, which is the recommended bounding value. Deposition velocity and residence time (half-life) of HTO in vegetation are related by the leaf area and leaf water volume in the forest. For the characteristics of the pine plantation at SRS the residence time corresponding to the average, daylight deposition velocity is 0.4 hours. The residence time corresponding to the night-time deposition velocity of 0.1 cm/s is around 2 hours. A simple dispersion model which accounts for deposition and re-emission of HTO vapor was used to evaluate the impact on exposure to the maximally exposed offsite individual (MOI) at the SRS boundary (Viner 2012). Under conditions that produce the bounding, 95 th percentile MOI exposure, i.e., low wind speed, weak turbulence, night, low deposition velocity, the effect of deposition and re-emission on MOI exposure was found to be very small. The exposure over the two hour period following arrival of the plume was found to be decreased by less than 0.05 %. Furthermore the sensitivity to deposition velocity was low. Increasing deposition velocity to 0.5 cm/s reduced exposure to 0.3 %. After a 24 hour period, an MOI would have been exposed to all of the released material. Based on the low sensitivity of MOI exposure to the value of deposition velocity when re-emission is considered, it is appropriately conservative to use a 0.0 cm/s effective deposition velocity for safety analysis in the MACCS2 code.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2012

Software Regression Quality Assurance for MACCS2 Version 2.5.0.0 through Version 2.5.0.9

The “State-of-the-Art Consequence Analyses (SOARCA) Project: Uncertainty Analysis” has used a newer version of the MELCOR Accident Consequence Code Systems Version 2 (MACCS2) than what was previously used in the “SOARCA Project Volume 1: Peach Bottom Integrated Analysis,” NUREG/CR-7110 Volume 1. A software regression quality assurance program has been implemented for MACCS2. The documentation for the quality assurance from MACCS2 Version 2.5.0.0 through 2.5.0.9 ensures continuity between code versions and provides insight into the changes made within MACCS2.

Bixler, Nathan E.↗

DOE OSTI2013

Analysis of Dose Consequences Arising from the Release of Spent Nuclear Fuel from Dry Storage Casks

The resulting dose consequences from releases of spent nuclear fuel (SNF) residing in a dry storage casks are examined parametrically. The dose consequences are characterized by developing dose versus distance curves using simplified bounding assumptions. The dispersion calculations are performed using the MELCOR Accident Consequence Code System (MACCS2) code. Constant weather and generic system parameters were chosen to ensure that the results in this report are comparable with each other and to determine the relative impact on dose of each variable. Actual analyses of site releases would need to accommodate local weather and geographic data. These calculations assume a range of fuel burnups, release fractions (RFs), three exposure scenarios (2 hrs and evacuate, 2 hrs and shelter, and 24 hrs exposure), two meteorological conditions (D-4 and F-2), and three release heights (ground level – 1 meter (m), 10 m, and 100 m). This information was developed to support a policy paper being developed by U.S. Nuclear Regulatory Commission (NRC) staff on an independent spent fuel storage installation (ISFSI) and monitored retrievable storage installation (MRS) security rulemaking.

Durbin, Samuel G.↗

DOE OSTI2018

SecPop Version 4: Sector Population Land Fraction and Economic Estimation Program: Users? Guide Model Manual and Verification Report.

In 1973 the U.S. Environmental Protection Agency (EPA) developed SecPop to calculate population estimates to support a study on air quality. The Nuclear Regulatory Commission (NRC) adopted this program to support siting reviews for nuclear power plant construction and license applications. Currently SecPop is used to prepare site data input files for offsite consequence calculations with the MELCOR Accident Consequence Code System (MACCS). SecPop enables the use of site-specific population, land use, and economic data for a polar grid defined by the user. Updated versions of SecPop have been released to use U.S. decennial census population data. SECPOP90 was released in 1997 to use 1990 population and economic data. SECPOP2000 was released in 2003 to use 2000 population data and 1997 economic data. This report describes the current code version, SecPop version 4.3, which uses 2010 population data and both 2007 and 2012 economic data. It is also compatible with 2000 census and 2002 economic data. At the time of this writing, the current version of SecPop is 4.3.0, and that version is described herein. This report contains guidance for the installation and use of the code as well as a description of the theory, models, and algorithms involved. This report contains appendices which describe the development of the 2010 census file, 2007 county file, and 2012 county file. Finally, an appendix is included that describes the validation assessments performed.

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2017

Development of Tritide Dose Assumptions for Tritium Processing and Extraction Facilities MACCS2 Consolidated Hazard Analysis and Documented Safety Analysis Revision

The purpose of this report is to identify the Tritide Dose Assumptions applicable to the Tritium Facilities. These dose assumptions support the MELCOR Accident Consequence Code System 2 (MACCS2) revision of the Consolidated Hazards Analysis (CHA) and Documented Safety Analysis (DSA) for the Tritium Facilities, including the Tritium Extraction Facility.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Development of Tritide Dose Assumptions for Tritium Processing and Extraction Facilities MACCS2 Consolidated Hazard Analysis and Documented Safety Analysis Revision

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2019

FGR 13 Dose Conversion Factor Files

The purpose of this document is to discuss the construction of two MELCOR Accident Consequence Code System (MACCS) dose conversion factor (DCF) files in some detail, an older file created in 2007 named FGR13DCF.inp and a newer file created in 2018 called FGR13GyEquiv_RevAinp. Very briefly, the difference between the two files is that the older file follows the standard conventions of assigning a radiation weighting factor of 20 for alpha radiation for all tissues and organs; whereas, the newer file complies with the Federal Guidance Report (FGR) 13 health effects modeling and uses modified radiation weighting factors (referred to as relative biological effectiveness (RBE) factors) for alpha radiation of 10 for breast and of 1 for red bone marrow. During an intermediate period, a file called FGR13GyEquiv.inp was created and used for the State-of-the-Art Reactor Consequence Analysis (SOARCA) calculations. This file was not released to the MACCS user community, but it is also discussed briefly in this document.

63 RADIATION, THERMAL, AND OTHER ENVIRON. POLLUTAN↗

DOE OSTI2020

Simplified Approach for Scoping Assessment of Non-LWR Source Terms

This report describes a structure to aid in evaluation of release mitigation strategies across a range of reactor technologies. The assessment performed for example reactor concepts utilizes previous studies of postulated accident sequences for each reactor concept. This simplified approach classifies release mitigation strategies based on a range of barriers, physical attenuation processes, and system performance. It is not, however, intended to develop quantitative estimates of radiological release magnitudes and compositions to the environment. Rather, this approach is intended to identify the characteristics of a reactor design concepts release mitigation strategies that are most important to different classes of accident scenarios. It uses a scoping methodology to provide an approximate, order-of-magnitude, estimate of the radiological release to the environment and associated off-site consequences. This scoping method is applied to different reactor concepts, considering the performance of barriers to fission product release for these concepts under sample accident scenarios. The accident scenarios and sensitivity evaluations are selected in this report to evaluate the role of different fission product barriers in ameliorating the source term to the environment and associated off-site consequences. This report applies this structure to characterize how release mitigation measures are integrated to define overall release mitigation strategies for High Temperature Gas Reactors (HTGRs), Sodium Fast Reactors (SFRs), and liquid fueled Molten Salt Reactors (MSRs). To support this evaluation framework, factors defining a chain of release attenuation stages, and thus an overall mitigation strategy, must be established through mechanistic source term calculations. This has typically required the application of an integral plant analysis code such as MELCOR. At present, there is insufficient evidence to support a priori evaluation of the effectiveness of a release mitigation strategy for advanced reactor concepts across the spectrum of events that could challenge the radiological containment function. While it is clear that these designs have significant margin to radiological release to the environment for the scenarios comprising the design basis, detailed studies have not yet been performed to assess the risk profile for these plants. Such studies would require extensive evaluation across a reasonably complete spectrum of accident scenarios that could lead to radiological release to the environment.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2020

AniMACCS User Guide

This SAND Report provides an overview of AniMACCS, the animation software developed for the MELCOR Accident Consequence Code System (MACCS). It details what users need to know in order to successfully generate animations from MACCS results, to include information on the capabilities, requirements, testing, limitations, input settings, and problem reporting instructions for AniMACCS version 1.3. Supporting information is provided in the appendices, such as guidance on required input files using both WinMACCS and running MACCS from the command line. This page left blank

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2021

Demonstration of MELCOR and MACCS Capabilities for Molten Salt Reactor Decay Heat Removal During both Normal Operations and Salt Spill Scenarios

This report provides a demonstration of MELCOR and MELCOR Accident Consequence Code System (MACCS) capabilities to perform a dose assessment for a Molten Salt Reactor (MSR) off-gas system. A primary containment system salt spill is used as the off-normal scenario, along with a normal operation dose assessment for comparison. This report discusses the tools, methods, and information used in this assessment so that it may be utilized as a starting point for future advanced reactor consequence analyses. This report also highlights several gaps, to include the need for reactor inventory information specific to advanced reactors, and the need for specific atmospheric transport models that take into account the unique deposition behaviors of tritium and carbon-14, and makes recommendations for closing these gaps. This report satisfies the DOE NE Milestone M4RD-21SN0601062.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗