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DOE OSTI2025

Direct Reactions and Spectroscopy with Hydrogen Targets at the RIBF

The implementation of MINOS, a liquid-hydrogen target and vertex-tracking system, at the Radioactive Isotope Beam Factory (RIBF) of the RIKEN Nishina Center has led to numerous nuclear physics discoveries. In experiments involving rare-isotope beams at intermediate energies, MINOS was integrated with the DALI2 + γ-ray spectrometer and other auxiliary devices. These were positioned in front of both the ZeroDegree Spectrometer and the SAMURAI magnetic spectrometers for the identification of ejectiles following quasi-free scattering direct reactions. Notable outcomes include the first spectroscopy of unstable doubly magic nucleus, investigations into multi-neutron decays, and the study of shell evolution along isotopic chains. These studies have significantly advanced our understanding of nuclear structure, as well as the mechanism of direct reactions with fast beams. Simultaneously, these direct reactions enabled raising new physics questions and paved the way for next-generation experimental setups to address them. Through the combination of particle- and γ-spectroscopy with a broad range of nuclear reactions, new frontiers will be established in our understanding of nuclear structure and reaction mechanisms. Here, we review the past ten years of studies on direct reactions and spectroscopy of exotic nuclei with liquid-hydrogen targets at the RIBF and define a vision for the future.

Bertulani, Carlos A [East Texas A & M Univ., Comme↗

DOE OSTI2014

Poster — Thur Eve — 41: Considerations for Patients with Permanently Implant Radioactive Sources Requiring Unrelated Surgery

Permanent implant of sealed radioactive sources is an effective technique for treating cancer. Typically, the radioactive sources are implanted in and near the disease, depositing dose locally over several months. There may be instances where these patients must undergo unrelated surgical procedures when the radioactive material remains active enough to pose risks. This work explores these risks, discusses strategies to mitigate those risks, and describes a case study for a permanent I-125 prostate brachytherapy implant patient who developed colo-rectal cancer and required surgery 6 months after brachytherapy. The first consideration is identifying the risk from unwarranted radiation to the patient and staff before, during, and after the surgical procedure. The second is identifying the risk the surgical procedure may have on the efficacy of the brachytherapy implant. Finally, there are considerations for controlling for radioactive substances from a regulatory perspective. After these risks are defined, strategies to mitigate those risks are considered. These strategies may include applying the concepts of ALARA, the use of protective equipment and developing a best practice strategy with the operating room team. We summarize this experience with some guidelines: If the surgical procedure is near (ex: 5 cm) of the implant; and, the surgical intervention may dislodge radioisotopes enough to compromise treatment or introduces radiation safety risks; and, the radioisotope has not sufficiently decayed to background levels; and, the surgery cannot be postponed, then a detailed analysis of risk is advised.

60 APPLIED LIFE SCIENCES↗

NASA NTRS2004

Very High Efficiency, Miniaturized, Long-Lived Alpha Particle Power Source Using Diamond Devices for Extreme Space Environments

A power source that converts a-particle energy into electricity by coulomb collision in doped diamond films is described. Alpha particle decay from curium-244 creates electron-hole pairs by free- ing electrons and holes inside the crystal lattice in N- and P-doped diamond films. Ohmic contacts provide electrical connection to an electronic device. Due to the built-in electric field at the rectifying junction across the hT- and P-doped diamond films, the free electrons are constrained to traveling in generally one direction. This one direction then supplies electrons in a manner similar to that of a battery. The radioactive curium layer may be disposed on diamond films for even distribution of a-particle radiation. The resulting power source may be mounted on a diamond substrate that serves to insulate structures below the diamond substrate from a-particle emission. Additional insulation or isolation may be provided in order to prevent damage from a-particle collision. N-doped silicon may be used instead of N-doped diamond.

Elizabeth A Kolawa↗

DOE OSTI2013

Radiological Impact Associated to Technologically Enhanced Naturally Occurring Radioactive Materials (TENORM) from Coal-Fired Power Plants Emissions - 13436

Certain materials used and produced in a wide range of non-nuclear industries contain enhanced activity concentrations of natural radionuclides. In particular, electricity production from coal is one of the major sources of increased human exposure to naturally occurring radioactive materials. A methodology was developed to assess the radiological impact due to natural radiation background. The developed research was applied to a specific case study, the Sines coal-fired power plant, located in the southwest coastline of Portugal. Gamma radiation measurements were carried out with two different instruments: a sodium iodide scintillation detector counter (SPP2 NF, Saphymo) and a gamma ray spectrometer with energy discrimination (Falcon 5000, Canberra). Two circular survey areas were defined within 20 km of the power plant. Forty relevant measurements points were established within the sampling area: 15 urban and 25 suburban locations. Additionally, ten more measurements points were defined, mostly at the 20-km area. The registered gamma radiation varies from 20 to 98.33 counts per seconds (c.p.s.) corresponding to an external gamma exposure rate variable between 87.70 and 431.19 nGy/h. The highest values were measured at locations near the power plant and those located in an area within the 6 and 20 km from the stacks. In situ gamma radiation measurements with energy discrimination identified natural emitting nuclides as well as their decay products (Pb-212, Pb-2142, Ra-226, Th-232, Ac-228, Th-234, Pa-234, U- 235, etc.). According to the results, an influence from the stacks emissions has been identified both qualitatively and quantitatively. The developed methodology accomplished the lack of data in what concerns to radiation rate in the vicinity of Sines coal-fired power plant and consequently the resulting exposure to the nearby population. (authors)

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2019

Defining Processing Times for Accelerator Produced 225Ac and Other Isotopes from Proton Irradiated Thorium

During the purification of radioisotopes, decay periods or time dependent purification steps may be required to achieve a certain level of radiopurity in the final product. Actinum-225 (Ac-225), Silver-111 (Ag-111), Astatine-211 (At-211), Ruthenium-105 (Ru-105), and Rhodium-105 (Rh-105) are produced in a high energy proton irradiated thorium target. Experimentally measured cross sections, along with MCNP6-generated cross sections, were used to determine the quantities of Ac-225, Ag-111, At-211, Ru-105, Rh-105, and other co-produced radioactive impurities produced in a proton irradiated thorium target at Brookhaven Linac Isotope Producer (BLIP). Ac-225 and Ag-111 can be produced with high radiopurity by the proton irradiation of a thorium target at BLIP.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2017

Air and radon pathways screenings methodologies for the next revision of the E-area PA

The strategic plan for the next E-Area Low-Level Waste Facility Performance Assessment includes recommended changes to the screening criteria used to reduce the number of radioisotopes that are to be considered in the air and radon pathways incorporated into the GoldSim® atmospheric release model (ARM). For the air pathway, a revised screening methodology was developed based on refinement of previous E-Area PA screening approaches and consideration of the strategic plan recommendations. The revised methodology has three sequential screening steps for each radioisotope: (1) volatility test using the Periodic Table of the Elements, (2) stability test based on half-life, and (3) stability test based on volatility as measured by the Henry’s Law constant for the assumed dominant gaseous species or vapor pressure in the case of tritiated water. Of the 1252 radioisotopes listed in the International Commission on Radiological Protection Publication 107, only the 10 that satisfied all three steps of the revised screening methodology will be included in the ARM. They are: Ar-37, Ar-39, Ar-42, C-14, H-3, Hg-194, Hg-203, Kr-81, Kr-85, and Xe-127. For the radon pathway, a revised screening methodology was developed that also has three sequential steps: (1) identify all decay chains that terminate at Rn-222, (2) screen out parents that decay through U-238 because of its 4.5-billion-year primordial half-life, and (3) eliminate remaining parents whose half-life is shorter than one day. Of the 86 possible decay chains leading to Rn-222, six decay chains consist of 15 unique radioisotopes that will be incorporated into the ARM. The 15 radioisotopes are: U-238, Th-234, Pa-234m, Pu-238, U-234, Th-230, Ra-226, Cf-246, Cm-242, Am-242m, Am-242, Np-238, Np-234, Pa-230, and Rn-222.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2021

Modeling Summary of ASNF in DOE Sealed Standard Canisters

A pathway for road-ready and final disposition packaging configurations for the aluminum-clad spent nuclear fuel (ANSF) fuel dictates storage within helium backfilled sealed DOE standard canisters. The typical packaging configuration for the 15-foot DOE canisters places 10 advance test reactor (ATR) elements within a Type 1a basket, and three baskets are loaded within each DOE canister. During in-reactor operations and cooling pond storage conditions, oxyhydroxide layers form on the surface of the aluminum clad fuel. These layers produce hydrogen gas over time due to the fuel’s radiation field. As part of the packing procedure, the ATR fuel should be dried to remove any residual physio-/chemi- sorbed water bound to the surface. The results of this modeling will include results at fully saturated and fully dried conditions. In addition, fuels that are currently stored at the Savannah River Site were also studied for their potential for hydrogen and pressure build up. These two additional fuels modeled were the Missouri University Research Reactor (MURR) fuel, which is packaged in the same configuration as the ATR, but with a 10-foot-tall DOE standard canister. This was selected due to its relatively high decay heat compared to other DOE-managed ASNF. The second additional fuel studied with the modeling effort was the High Flux Isotope Reactor (HFIR) fuel. This fuel is modeled as two separate DOE canisters with the inner and outer annulus split for storage. The HFIR was selected for study due to its high aluminum cladding surface area. In the associated experimental work, Task 2 - Oxyhydroxide Layer Radiolytic Gas Generation Resolution, additional tests were completed in a helium environment, and updated G-values for the radiolytic production of hydrogen from the oxyhydroxide layers were provided. These values were 2.92 ×10 -4 µmol/J at 50% relative humidity and 4.12 ×10 -4 µmol/J at 100% relative humidity. These values are lower than the value for Argon that was used in prior modeling results. In addition, prior modeling results have been completed with the G-value applied to just the mass of the corrosion layer, and this has been updated to apply the G-value for hydrogen generation to the full mass of the fuel. These two effects combine to show much smaller pressure and hydrogen build up for the sealed canister model. For a nominal scenario of stored ATR fuel, after 50 years the model results give a 1.36 atm total pressure, 7% mole percent hydrogen, for the upper decay heat, 1.51 atm total pressure, 16% mole percent hydrogen, and for upper decay heat with undried fuel 2.6 atm total pressure, 15% mole percent hydrogen. For the MURR nominal case, the model results give 1.34 atm and 6% hydrogen, for upper decay heat this gives 1.41 atm total pressure with 10.8 % hydrogen, and for upper decay heat with undried fuel, this gives 2.38 atm total pressure with 9.9% hydrogen. The nominal scenario for HFIR fuel gives 1.39 atm total pressure with 9.9% hydrogen, the upper decay heat case gives 1.43 atm with 12.1% hydrogen, and the upper decay heat with undried fuel gives 2.17 atm total pressure with 11.9% hydrogen. These results confirm the ATR scenario bounds the other intact ASNF modeled here. No case modeled yields significant oxygen, and the lower decay heat cases for all fuels modeled have hydrogen concentrations that are under the 4% flammability limit after 50 years of storage. In addition, the modeled pressures for all cases are all significantly below the 500-psi limit for the DOE standard sealed canister.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2014

The light curve of SN 1987A revisited: constraining production masses of radioactive nuclides

We revisit the evidence for the contribution of the long-lived radioactive nuclides {sup 44}Ti, {sup 55}Fe, {sup 56}Co, {sup 57}Co, and {sup 60}Co to the UVOIR light curve of SN 1987A. We show that the V-band luminosity constitutes a roughly constant fraction of the bolometric luminosity between 900 and 1900 days, and we obtain an approximate bolometric light curve out to 4334 days by scaling the late time V-band data by a constant factor where no bolometric light curve data is available. Considering the five most relevant decay chains starting at {sup 44}Ti, {sup 55}Co, {sup 56}Ni, {sup 57}Ni, and {sup 60}Co, we perform a least squares fit to the constructed composite bolometric light curve. For the nickel isotopes, we obtain best fit values of M({sup 56}Ni) = (7.1 ± 0.3) × 10{sup –2} M {sub ☉} and M({sup 57}Ni) = (4.1 ± 1.8) × 10{sup –3} M {sub ☉}. Our best fit {sup 44}Ti mass is M({sup 44}Ti) = (0.55 ± 0.17) × 10{sup –4} M {sub ☉}, which is in disagreement with the much higher (3.1 ± 0.8) × 10{sup –4} M {sub ☉} recently derived from INTEGRAL observations. The associated uncertainties far exceed the best fit values for {sup 55}Co and {sup 60}Co and, as a result, we only give upper limits on the production masses of M({sup 55}Co) < 7.2 × 10{sup –3} M {sub ☉} and M({sup 60}Co) < 1.7 × 10{sup –4} M {sub ☉}. Furthermore, we find that the leptonic channels in the decay of {sup 57}Co (internal conversion and Auger electrons) are a significant contribution and constitute up to 15.5% of the total luminosity. Consideration of the kinetic energy of these electrons is essential in lowering our best fit nickel isotope production ratio to [{sup 57}Ni/{sup 56}Ni] = 2.5 ± 1.1, which is still somewhat high but is in agreement with gamma-ray observations and model predictions.

79 ASTRONOMY AND ASTROPHYSICS↗

DOE OSTI2013

Functional Bounding Content Envelope for Gamma and Neutron Emitting Isotopes in the Model 9977 Packaging

The contents approved for shipment in a Type B radioactive material transportation package have historically been descriptions of discrete items, or groupings of well-defined similar items, in the package safety basis documentation. The need for a comprehensive functional content envelope of both gamma and neutron emitting nuclides compliant with regulatory limits has become necessary as the DOE complex requires shipments of unique mixtures of radionuclides and impurities. Recent publications have presented a calculational model and a corresponding content envelope intended to be compliant with federal regulatory external radiation limits as well as design decay heat limits based on the Model 9977 Packaging. The methodology used to develop this content envelope consisted of determining the external radiation dose rates based on one gram of a given isotope combined, in the case of actinides, with various levels of light element impurities and determining the allowable mass to the regulatory limits based on these dose rates. The method of ratioing from these calculations to the masses that meet the regulatory limits fails in the case of some combinations of actinides and light element impurities because of the effect of subcritical multiplication, resulting in non-conservative and non-compliant dose rates in some cases. This is particularly true for many actinides combined with beryllium, boron, fluorine, lithium, and sodium. In this study, the source was modeled as a sphere with the appropriate dimensions based on the actinide or gamma source density and placed at the bottom of the containment vessel for maximum conservatism. The results presented in this report rectify the above-noted deficiencies by adjusting the previously determined mass limits so that the resulting dose rates are compliant with regulatory limits. Some amount of iteration was required in some instances to adjust the masses for dose compliance. In some instances the masses were increased since they were too conservative. In addition, the method of ratioing the masses when applied to several gamma emitting isotopes also resulted in either non-conservative or overly conservative data. The mass imits presented here are for bare sources and are complaint with the design decay heat limit of 19W. The masses are based on package surface dose rates (the limiting dose rate) that fall in the range 185-195 mrem/h, thus giving an additional margin of between 7.5% and 2.5% to the regulatory limit of 200 mrem/h. This report presents a revised set of mass limits (i.e., content envelope) for several neutron emitting actinides with varying levels of light element impurities compliant with both external radiation and design decay heat limits. In addition, revised limits for gamma sources are also presented. The neutron emitter mass limits are in Tables 1 through 9 and the gamma emitter mass limits are in Table 10. The revised content envelope covering a wide range of materials present in the DOE complex, was developed for the Model 9977 Packaging, providing conservative limits for other commonly-used Type B Packagings such as the Models 9975 and 9978. It should be noted that for the Model 9977 package, shipment is also possible with three shielded containers – lead or tungsten for gamma sources and polyethylene for neutron emitters. The design decay heat limits for these are as follows: 19W for tungsten, 6W for lead and 3W for polyethylene. The limits presented in this report would need to be checked against the lower decay heat limits for the lead and polyethylene and adjusted as required for compliance. Criticality safety is not factored into these limits and should be evaluated on a case by case basis.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2014

Metastable Radioxenon Verification Laboratory (MRVL) Year-End Report

This is the year end report that is due to the client. The MRVL system is designed to measure multiple radioxenon isotopes ( 135 Xe, 133 Xe, 133 mXe and 133 mXe) simultaneously. The system has 12 channels to load samples and make nuclear measurements. Although the MRVL system has demonstrated excellent stability in measurements of Xe-133 and Xe-135 over the year of evaluation prior to delivery, there has been concern about system stability over measurements performed on samples with orders of magnitude different radioactivity, and samples containing multiple isotopes. To address these concerns, a series of evaluation test have been performed at the end-user laboratory. The evaluation was performed in two separate phases. Phase 1 made measurements on isotopically pure Xe-133 from high radioactivity down to the system background levels of activity, addressing the potential count rate dependencies when activities change from extreme high to very low. The second phase performed measurements on samples containing multiple isotopes (Xe-135, Xe-133 and Xe-133m), and addressed concerns about the dependence of isotopic concentrations on the presence of additional isotopes. The MRVL showed a concentration dependence on the Xe-133 due to the amount of Xe-133m that was in the sample. The dependency is due to the decay of Xe-133m into Xe-133. This document focuses on the second phase and will address the analysis used to account for ingrowth of Xe-133 from Xe-133m.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2019

Integrated Approach to Fluoride High Temperature Reactor Technology and Licensing Challenges (FHR-IRP)

This document constitutes progress made on this project by the Georgia Tech led Integrated Approach to Fluoride High Temperature Reactor Technology and Licensing Challenges (FHR-IRP). This is the first draft of the project final report. Chapter 3 is focused on Tritium management. Tritium management in Fluoride-salt High-temperature Reactors (FHRs) presents a major operational concern and a significant challenge. Tritium is generated via multiple pathways including various interactions in fuel and structural materials, as well as coolant (primarily interaction with lithium and beryllium), leading to a complex evolution and spatial distribution of tritium generation rate. FHR generates orders of magnitude more tritium than in LWRs, and has potential for a significant release, via permeation through structural materials (due to high temperature), and through heat exchanger(s). Therefore, accurate estimate is necessary. For that purpose relevant tritium production pathways were identified. A methodology to obtain space-time dependent tritium generation rate was developed and implemented, employing SCALE and MCNP codes. Detailed models of the reference Advanced High Temperature Reactor (AHTR) design were developed. A multi-parametric metrics was introduced to capture relevant tritium production indicators, including the instantaneous tritium production rate, total tritium production (neglecting subsequent decay), and accumulated tritium production (accounting for the decay). The obtained results will inform tritium management, including the sizing of the components, and to gain insight on the impact of the core design parameters on tritium generation rates. Research effort on Liquid Salt Impurity Removal, Redox and Corrosion Control is discussed in Chapter 4. Section 4.1 is an introduction chapter. Based on the current available knowledge and operation experience on previous MSRs and other types of nuclear reactors, the report identifies the major impurities that are related to either material compatibility, salt properties or radioactive source terms, and the potential contamination sources of these impurities, their transport in the primary coolant loop and available cleanup methodologies for a particular group of impurity. In section 4.2, the available purification methods were discussed, including the liquid/liquid extraction using liquid Bi-Li which has been applied in MSBR and electrochemical purification which is developed based on Pyroprocessing. In this section, the new multi-stage and multi-salt-component was developed and discussed. Section 4.3 discussed the redox potential control for salt chemistry and corrosion mitigation, including the control theory and the soluble/soluble buffer control method. The section also developed method on how to analyze soluble/soluble redox reaction in a molten salt system and designed an electrochemical sensor for redox potential control. Corrosion models were also reported in this section. Section 4.4 focused on how to measure the fundamental properties in a molten salt. As an example, the section focused on La fluoride in the molten salt. The section developed methods on how analyze electrochemical measurement data to obtain the fundamental properties including the redox potential, diffusion coefficient and exchange current density. All these fundamental information is important for salt purification system development. Section 4.5 focused on measurements of corrosion products in the salts. Because of the different behaviors between Cr and La in the salt, new analysis method for Cr fluorides in the salt was developed. Some results for Fe fluoride in the salt and corrosion measurements were also reported. Section 4.6 demonstrated the separation of oxide from the salt. This section also measured the solubility of La 2 O 3 in the salt as function of temperature. Section 4.7 focused on the separation of La and Ce from the molten salt. These two elements were selected because La was found to be separated by using inert electrode, while for separation of Ce, reactive electrode must be selected. Section 4.8 demonstrated the separation iodine ions from the molten salt. Two kinds of working electrode materials (metal W and graphite) were investigated. Section 4.9 focused on the fundamental data of fission products and actinides in liquid Bi. The chapter reviewed the available data, then based a CALPHAD model, the available data were broadened to the concentration-dependent cases.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

Equilibrium Temperature Profiles within Fission Product Waste Forms

We studied waste form strategies for advanced fuel cycle schemes. Several options were considered for three waste streams with the following fission products: cesium and strontium, transition metals, and lanthanides. These three waste streams may be combined or disposed separately. The decay of several isotopes will generate heat that must be accommodated by the waste form, and this heat will affect the waste loadings. To help make an informed decision on the best option, we present computational data on the equilibrium temperature of glass waste forms containing a combination of these three streams.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2025

Spectral line-shape in collinear laser spectroscopy after atomic charge exchange

Collinear laser spectroscopy experiments on fast, neutral beams have been extensively used for studies on short-lived radioactive nuclei, taking advantage of its high sensitivity. The resulting resonance line-shape is known to show significant distortion, due to the energy exchange during the charge-exchange neutralization process, which can cause large systematic uncertainty in the determined centroid. A model for the line shape was constructed and simulated to be compared to measured Al, Si, and Ni hyperfine spectra. It is shown that the distortion is caused mainly by the transfer of electron into many different energy levels in the projectile atom and subsequent decays, rather than secondary inelastic collisions, which were often assumed in the line shape analysis before. Furthermore, the model can also be applied to other projectile–alkali pairs, providing a reliable line-shape with less fitting parameters than conventional phenomenological models.

Atom↗

DOE OSTI2018

Classification of Reactor Facility Operational State Using SPRT Methods with Radiation Sensor Networks

We consider the problem of inferring the operational state of a reactor facility by using measurements from a radiation sensor network, which is deployed around the facility’s ventilation stack. The radiation emissions from the stack decay with distance, and the corresponding measurements are inherently random with parameters determined by radiation intensity levels at the sensor locations. We fuse measurements from network sensors to estimate the intensity at the stack, and use this estimate in a one-sided Sequential Probability Ratio Test (SPRT) to infer the on/off state of the reactor facility. We demonstrate the superior performance of this method over conventional majority vote fusers and individual sensors using (i) test measurements from a network of NaI sensors, and (ii) emulated measurements using radioactive effluents collected at a reactor facility stack. We analytically quantify the performance improvements of individual sensors and their networks with adaptive thresholds over those with fixed ones, by using the packing number of the radiation intensity space.

Ramirez Aviles, Camila A.↗

DOE OSTI2018

Advances in Geologic Disposal Safety Assessment and an Unsaturated Alluvium Reference Case

The Spent Fuel and Waste Science and Technology (SFWST) Campaign of the U.S. Depat ment of Energy (DOE) Office of Nuclear Energy (NE), Office of Fuel Cycle Technology (OFCT) is conducting research and development (R&D) on geologic disposal of spent nuclear fuel (SNF) and high level nuclear waste (HLW). Two high priorities for SFWST disposal R&D are design concept development and disposal system modeling (DOE 2011, Table 6). These priorities are directly addressed in the SFWST Geologic Disposal Safety Assessment (GDSA) work package, which is charged with developing a disposal system modeling and analysis capability for evaluating disposal system performance for nuclear waste in geologic media. This report describes specific GDSA activities in fiscal year 2018 (FY 2018) toward the development of GDSA Framework, an enhanced disposal system modeling and analysis capability for geologic disposal of nuclear waste. GDSA Framework employs the PFLOTRAN thermal-hydrologic-chemical multiphysics code (Hammond et al. 2011a; Lichtner and Hammond 2012) and the Dakota uncertainty sampling and propagation code (Adams et al. 2012; Adams et al. 2013). Each code is designed for massivelyparallel processing in a high-performance computing (HPC) environment. Multi-physics representations in PFLOTRAN are used to simulate various coupled processes including heat flow, fluid flow, waste dissolution, radionuclide release, radionuclide decay and ingrowth, precipitation and dissolution of secondary phases, and radionuclide transport through engineered barriers and natural geologic barriers to the biosphere. Dakota is used to generate sets of representative realizations and to analyze parameter sensitivity.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2016

SCALE Code System 6.2.1

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

SCALE Code System 6.2.2

The SCALE Code System is a widely used modeling and simulation suite for nuclear safety analysis and design that is developed, maintained, tested, and managed by the Reactor and Nuclear Systems Division (RNSD) of Oak Ridge National Laboratory (ORNL). SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor physics, radiation shielding, radioactive source term characterization, and sensitivity and uncertainty analysis. Since 1980, regulators, licensees, and research institutions around the world have used SCALE for safety analysis and design. SCALE provides an integrated framework with dozens of computational modules including 3 deterministic and 3 Monte Carlo radiation transport solvers that are selected based on the desired solution strategy. SCALE includes current nuclear data libraries and problem-dependent processing tools for continuous-energy (CE) and multigroup (MG) neutronics and coupled neutron-gamma calculations, as well as activation, depletion, and decay calculations. SCALE includes unique capabilities for automated variance reduction for shielding calculations, as well as sensitivity and uncertainty analysis. SCALE’s graphical user interfaces assist with accurate system modeling, visualization of nuclear data, and convenient access to desired results. SCALE 6.2 represents one of the most comprehensive revisions in the history of SCALE, providing several new capabilities and significant improvements in many existing features.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

SCALE Code System

The SCALE Code System is a widely-used modeling and simulation suite for nuclear safety analysis and design that is developed, maintained, tested, and managed by the Reactor and Nuclear Systems Division (RNSD) of Oak Ridge National Laboratory (ORNL). SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor and lattice physics, radiation shielding, spent fuel and radioactive source term characterization, and sensitivity and uncertainty analysis. Since 1980, regulators, licensees, and research institutions around the world have used SCALE for safety analysis and design. SCALE provides an integrated framework with dozens of computational modules including three deterministic and three Monte Carlo radiation transport solvers that are selected based on the desired solution strategy. SCALE includes current nuclear data libraries and problem-dependent processing tools for continuous-energy (CE) and multigroup (MG) neutronics and coupled neutron-gamma calculations, as well as activation, depletion, and decay calculations. SCALE includes unique capabilities for automated variance reduction for shielding calculations, as well as sensitivity and uncertainty analysis. SCALE’s graphical user interfaces assist with accurate system modeling, visualization of nuclear data, and convenient access to desired results.SCALE 6.2 provides many new capabilities and significant improvements of existing features.New capabilities include:• ENDF/B-VII.1 nuclear data libraries CE and MG with enhanced group structures,• Neutron covariance data based on ENDF/B-VII.1 and supplemented with ORNL data,• Covariance data for fission product yields and decay constants,• Stochastic uncertainty and correlation quantification for any SCALE sequence with Sampler,• Parallel calculations with KENO,• Problem-dependent temperature corrections for CE calculations,• CE shielding and criticality accident alarm system analysis with MAVRIC,• CE depletion with TRITON (T5-DEPL/T6-DEPL),• CE sensitivity/uncertainty analysis with TSUNAMI-3D,• Simplified and efficient LWR lattice physics with Polaris,• Large scale detailed spent fuel characterization with ORIGAMI and ORIGAMI Automator,• Advanced fission source convergence acceleration capabilities with Sourcerer,• Nuclear data library generation with AMPX, and• Integrated user interface with Fulcrum.Enhanced capabilities include:• Accurate and efficient CE Monte Carlo methods for eigenvalue and fixed source calculations,• Improved MG resonance self-shielding methodologies and data,• Resonance self-shielding with modernized and efficient XSProc integrated into most sequences,• Accelerated calculations with TRITON/NEWT (generally 4x faster than SCALE 6.1),• Spent fuel characterization with 1470 new reactor-specific libraries for ORIGEN,• Modernization of ORIGEN (Chebyshev Rational Approximation Method [CRAM] solver, API for high-performance depletion, new keyword input format)• Extension of the maximum mixture number to values well beyond the previous limit of 2147 to ~2 billion,• Nuclear data formats enabling the use of more than 999 energy groups,• Updated standard composition library to provide more accurate use of natural abundances, andvi• Numerous other enhancements for improved usability and stability.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗