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All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Second Target Station (STS) Project: Analysis of Bunker Shielding–Preliminary Design

At the Department of Energy’s Oak Ridge National Laboratory, the Second Target Station (STS) beamline sources for preliminary design have been used to perform a shielding analysis of the bunker. The simulations in this report used the ST11 beamline source, primarily the ST11 neutron source, because that source provides conservative prompt total dose rate estimates for all beamlines. Prompt total dose rates (i.e., neutron plus photon dose rates when the beam is on) were calculated on top of the bunker roof, outside the bunker wall, and in the basement below the bunker and instrument floor. Design changes were made to ensure the prompt total dose rate in these areas does not exceed 0.25 mrem/h. This report documents the required shielding thicknesses to meet this dose rate limit. There is one exception to this: for a combination of open and closed beamlines, the prompt total dose rate outside the bunker across from the closed beamline, where there is less shielding due to the lack of beamline shielding, slightly exceeds 0.25 mrem/h. A future analysis will document the shielding modifications required to reduce the calculated prompt total dose rates for this configuration to less than 0.25 mrem/h once more details regarding the STS high density concrete density and composition are known. Finally, prompt total dose rates inside the bunker were simulated. The results of this simulation show that the majority of prompt total dose rate in the bunker is due to neutrons, and a few percent of the total is caused by secondary photons created by the neutrons. The contribution to prompt total dose rate inside the bunker by the primary photons that enter each beamline from the target/moderator area is at least two orders of magnitude less than the neutron contribution. This means prompt total dose rate contributions due to primary photons inside and outside the bunker can be neglected.

42 ENGINEERING↗
DOE OSTI2025

2025 Review and Revision of Federal Guidance Report 15

Federal Guidance Report No. 15 (FGR 15), External Exposure to Radionuclides in Air, Water and Soil, published in 2019 and referred to below as FGR 2019, provides age-specific effective dose rate coefficients for reference persons externally exposed to each of 1252 radionuclides homogenously distributed in environmental media. Soon after completion of FGR 2019, the International Commission on Radiological Protection (ICRP) published a similar report (ICRP Publication 144, 2020) addressing the same radionuclides and many of the external exposure scenarios addressed in FGR 2019. In 2023 Argonne National Laboratory (ANL) published a review of dose coefficients in FGR 2019, concluding that “The external effective dose from beta radiation is not appropriately accounted for in FGR 15 for both low- and high-energy beta emitters in air, in soil, and on soil surfaces.” That conclusion was based largely on comparisons of dose coefficients in FGR 2019 with values in ICRP Publication 144 but also on consideration of some unexpected patterns of equivalent dose rate coefficients across tissues and of effective dose rate coefficients across different soil depths, for a selected set of low-energy beta emitters. In response to the ANL report, the Center for Radiation Protection Knowledge (CRPK) at Oak Ridge National Laboratory (ORNL) performed an extensive reexamination of the methods and published values of FGR 2019. CRPK found that coding errors had resulted in inaccurate estimates, primarily overestimates, in many of the dose coefficients tabulated in FGR 2019 but that many of the differences between results in FGR 2019 and ICRP Publication 144 could be traced to differences in methodology. CRPK has corrected all errors in FGR 2019 associated with the coding errors and has taken the opportunity to improve a major portion of the remaining dose coefficients in FGR 2019, primarily through additional Monte Carlo calculations resulting in improved statistics for tissue equivalent dose rate coefficients for exposure to monoenergetic sources. This has eliminated the need for extrapolation of results from high and medium energies to low energies, as was done in FGR 2019. In addition, inconsistencies in the methodology identified in the review, such as computational representation of the adult female, were eliminated. The revised effective dose rate coefficients are consistent with values in ICRP Publication 144 except for differences in values clearly arising from differences in methodology; and differences in values for a relatively small set of very low-energy radionuclides with highly uncertain dose coefficients regardless of methodology.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2020

Intensity modulated Ir-192 brachytherapy using high-Z 3D printed applicators

Gynecologic cancers are often asymmetric, yet current Ir-192 brachytherapy techniques provide only limited radial modulation of the dose. The shielded solutions investigated here solve this by providing the ability to modulate between highly asymmetric and radially symmetric dose distributions at a given location. To find applicator designs that can modulate between full dose and less than 50% dose, at the dimensions of the urethra, a 2D calculation algorithm was developed to narrow down the search space. Two shielding design types were then further investigated using Monte Carlo and Boltzmann-solver dose calculation algorithms. 3D printing techniques using ISO 10993 certified biocompatible plastics and 3D printable tungsten-loaded plastics were tested. It was also found that shadowing effects set by the shape of the shielding cannot be easily modulated out, hence careful design is required. The shielded applicator designs investigated here, allow for reduction of the dose by over 50% at 5 mm from the applicator surface in desired regions, while also allowing radially symmetric dose with isodose line deviations less than 0.5 mm from circular. The shielding designs were also chosen with treatment delivery time in mind. Treatment times for these shielded designs were found to be less than 1.4 times longer than a 6-channel unshielded cylinder for the equivalent fully symmetric dose distribution. The 2D calculation methods developed here provide a simple way to rapidly evaluate shielding designs, while the 3D printing techniques also allow for devices with novel shapes to be rapidly prototyped. Both TOPAS Monte Carlo and Acuros BV calculations show that significant dose shaping and organ at risk sparing can be achieved without significantly compromising the plan in regions that require the full dose.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2021

Pre- and post-treatment image-based dosimetry in 90 Y-microsphere radioembolization using the TOPAS Monte Carlo toolkit

Objective.To evaluate the pre-treatment and post-treatment imaging-based dosimetry of patients treated with 90Y-microspheres, including accurate estimations of dose to tumor, healthy liver and lung. To do so, the Monte Carlo (MC) TOPAS platform is in this work extended towards its utilization in radionuclide therapy. Approach. In this study, five patients treated at the Massachusetts General Hospital were selected for this study. All patients had data for both pre-treatment SPECT-CT imaging using 99mTc-MAA as a surrogate of the 90Y-microspheres treatment and SPECT-CT imaging immediately after the 90Y activity administration. Pre- and post-treatment doses were computed with TOPAS using the SPECT images to localize the source positions and the CT images to account for tissue inhomoegeneities. We compared our results with analytical calculations following the voxel-based MIRD scheme. Main results. TOPAS results largely agreed with the MIRD-based calculations in soft tissue regions: the average difference in mean dose to the liver was 0.14 Gy GBq -1 (2.6%). However, dose distributions in the lung differed considerably: absolute differences in mean doses to the lung ranged from 1.2 to 6.3 Gy GBq -1 and relative differences from 153% to 231%. We also found large differences in the intra-hepatic dose distributions between pre- and post-treatment imaging, but only limited differences in the pulmonary dose. Significance. Doses to lung were found to be higher using TOPAS with respect to analytical calculations which may significantly underestimate dose to the lung, suggesting the use of MC methods for 90Y dosimetry. According to our results, pre-treatment imaging may still be representative of dose to lung in these treatments.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2021

National Emission Standards for Hazardous Air Pollutants – Radionuclide Emissions Calendar Year 2020

The U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Field Office (NNSA/NFO) operates the Nevada National Security Site (NNSS) and the North Las Vegas Facility (NLVF). From 1951 through 1992, the NNSS was the continental testing location for U.S. nuclear weapons. Radionuclides in air from NNSS activities have been monitored since the initiation of atmospheric testing. After 1962, testing was limited to underground detonations, which greatly reduced radiation exposure to the public. Since the end of nuclear testing in 1992, radiation monitoring has focused on detecting airborne radionuclides from historically contaminated soils because this sources dominates the potential offsite dose. These radionuclides are derived from re-suspension of soil (primarily by wind) and emission of tritium-contaminated soil moisture through evapotranspiration. Low amounts of legacy-related tritium are also emitted to air at the NLVF, an NNSS support complex in North Las Vegas. To protect the public from harmful levels of manmade radiation, the Clean Air Act, National Emission Standards for Hazardous Air Pollutants (NESHAP), specifically the National Emission Standards for Emissions of Radionuclides Other Than Radon From Department of Energy Facilities (40 CFR 61, Subpart H, 2020) limits the release of radioactivity from a DOE facility to that which would cause 10 millirem per year (mrem/y) effective dose equivalent (EDE) to any member of the public. This limit does not include radiation unrelated to NNSS activities. Unrelated doses could come from naturally occurring radioactive elements, from sources such as medically or commercially used radionuclides, or from sources outside of the United States, such as Japan’s Fukushima nuclear power plant, which was damaged in 2011. NNSA/NFO demonstrates compliance with the NESHAP limit by reporting environmental measurements of radionuclide air concentrations at critical receptor locations on the NNSS. This alternative was proposed and formerly submitted to the U.S. Environmental Protection Agency (EPA) in 2001 (EPA 2001a) and has been the method used to demonstrate compliance with the 40 CFR 61.92 dose standard since 2005. Six locations on the NNSS have been established to act as critical receptor locations to demonstrate compliance with the NESHAP limit. These locations are closer to radionuclide releases than where the public resides so they act as protective substitutes for public receptor locations. Compliance is demonstrated if the measured annual average concentration is less than the NESHAP Concentration Level (CL) for Environmental Compliance listed in Table 2 of 40 CFR 61, Appendix E. For multiple radionuclides, compliance is demonstrated when the sum of the fractions (determined by dividing each radionuclide’s concentration by its CL and then adding the fractions together) is less than 1.0. The EPAapproved air transport model, called the Clean Air Package 1988 (CAP88-PC) is also used to calculate the effective dose equivalent to the maximally exposed individual from NNSS air emissions. CAP88-PC was also used to calculate the population dose, or the collective EDE (expressed as person-rem [roentgen equivalent man] per year [person-rem/y]) for all individuals combined who reside within 80 kilometers (km) of NNSS emission sources. In 2020, the potential dose from radiological emissions to air from both current and past NNSS activities was well below the 10 mrem/y dose limit. This is demonstrated by both the air sampling data collected at critical receptor air monitoring stations and CAP88-PC modeling. The average concentrations of radioactivity at air critical receptor stations ranged from 0.2% to a maximum of 4.2% of the allowed NESHAP limit. CAP88-PC modeling of all 2020 NNSS radionuclide emissions showed the maximally exposed individual to be in Amargosa Valley and this individual received a potential dose of 0.063 mrem/y. The collective dose was calculated to be 0.29 person-rem/year for the 521,300 people who lived within 80 km of NNSS emission sources.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2021

Use of Multi-Site Radiation Therapy for Systemic Disease Control

Metastatic cancer is a heterogeneous entity, some of which could benefit from local consolidative radiation therapy (RT). Although randomized evidence is growing in support of using RT for oligometastatic disease, a highly active area of investigation relates to whether RT could benefit patients with polymetastatic disease. This article highlights the preclinical and clinical rationale for using RT for polymetastatic disease, proposes an exploratory framework for selecting patients best suited for these types of treatments, and briefly reviews potential challenges. The goal of this hypothesis-generating review is to address personalized multimodality systemic treatment for patients with metastatic cancer. The rationale for using high-dose RT is primarily for local control and immune activation in either oligometastatic or polymetastatic disease. However, the primary application of low-dose RT is to activate distinct antitumor immune pathways and modulate the tumor stroma in efforts to better facilitate T cell infiltration. We explore clinical cases involving high- and low-dose RT to demonstrate the potential efficacy of such treatment. We then group patients by extent of disease burden to implement high- and/or low-dose RT. Patients with low-volume disease may receive high-dose RT to all sites as part of an oligometastatic paradigm. Subjects with high-volume disease (for whom standard of care remains palliative RT only) could be treated with a combination of high-dose RT to a few sites for immune activation, while receiving low-dose RT to several remaining lesions to enhance systemic responses from high-dose RT and immunotherapy. We further discuss how emerging but speculative concepts such as immune function may be integrated into this approach and examine therapies currently under investigation that may help address immune deficiencies. The review concludes by addressing challenges in using RT for polymetastatic disease, such as concerns about treatment planning workflows, treatment times, dose constraints for multiple-isocenter treatments, and economic considerations.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2020

Neutron heating assessment in Minerve reactor by using alanine/ESR dosimetry

In the framework of the ICERR (International Center based on Research Reactors) project, the effect of neutron/gamma irradiation on material and dosimeters has been studied. In this work, we explore the alanine dose-response in the core of the MINERVE research reactor at the CEA Cadarache and we investigate the neutron heating effects in order to evaluate alanine neutron sensitivity. The measured dose of the alanine detector in the mixed gamma-neutron field was estimated by electron spin resonance spectrometry (ESR) with reference to an absorbed dose in water from a 60 Co gamma-ray beam. The experimental sources of uncertainties were reduced by using the optimum conditions of charged-particle equilibrium (CPE), obtained by introducing alanine dosimeters into the aluminum pillbox. Cavity correction factors and dose components were evaluated by the Monte Carlo code MCNP. The calculation showed an underestimated neutron dose with a (C/E) value of about 0.993 +/- 10.83%(k = 1). Results indicate that the fast neutron dose, mainly due to elastic neutron scattering on hydrogen nuclei, represents about 70% of the total dose, while the gamma dose was found to be around 45% of the neutron dose. Additionally, the neutron relative efficiency (REn) to 60 Co gamma-radiation was estimated to be 0.544 +/- 8%(k = 1).

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Evolution of {delta}ferrite in a CF3 cast stainless steel upon neutron irradiation to 3, 5, 10, 20, and 40 dpa

The microstructural evolution of delta ferrite in a CF3 cast stainless steel irradiated to 3, 5, 10, 20, and 40 dpa with two dose rates was studied with atom probe tomography (APT) and transmission electron mi-croscopy. Spinodal decomposition and G-phase precipitates induced by neutron irradiation and thermal aging were quantified systematically. The neutron irradiation significantly enhances the spinodal decom-position in delta ferrite as both the spinodal wavelength (i.e. a characteristic repeat distance) and am-plitude (i.e. magnitude of elemental concentration fluctuation) increase after irradiation. The wavelength only varies slightly when the dose increases from 5 to 40 dpa, while the amplitude increases dramatically from 10 to 20 dpa and starts to saturate before 20 dpa. The dose rate and irradiation temperature also have a notable effect on the spinodal wavelength and amplitude. The study shows that higher dose rate promotes a larger wavelength at given irradiation doses. Regarding the G-phase precipitates, a slightly lower irradiation temperature would result in a smaller mean size of G-phase precipitates, while a lower irradiation dose rate would lead to a larger G-phase precipitates at a given temperature and dose. Over-all, the spinodal decomposition and G-phase precipitates in the delta ferrite continue to evolve with the increasing dose beyond 10 dpa. This study confirms that the formation of G-phase precipitates at the in-terdomain region between alpha and alpha-prime is facilitated by the Si and Mo atoms rejected from Fe rich alpha phase and the Ni and Mn atoms rejected from Cr rich alpha-prime phase. Neutron irradiation plays a dominant role in the ferrite instability, and the effect of prior thermal aging at 400 degrees C for 10,000 hours is negligible as the dose is 3 dpa and above.

duplex stainless steel↗
DOE OSTI2021

FY2020 Savannah River Site Composite Analysis Annual Summary Review

This document provides the Department of Energy (DOE) Order 435.1 and its manual, Radioactive Waste Management (DOE 2021a, 2021b) required Annual Review for the Savannah River Site (SRS) Composite Analysis (CA). Progress made to-date toward addressing the secondary issue from the LFRG review of the 2010 SRS CA has focused primarily upon inventory estimate improvements. Inventory impacts dose in a linear fashion and reduces the uncertainty with the CA conclusions. Maintenance items are addressed, as funding allows, based on the relative risk associated with meeting the performance objectives. Currently, there is minimal risk in exceeding the DOE 100 mrem/yr CA primary dose limit or the DOE 30 mrem/yr dose constraint (administrative limit). Proposed activities, discoveries, new information and changes potentially affecting the 2010 SRS CA are documented in this and earlier Annual Summary reports, and a consolidated list of changes since the 2010 CA is documented in this report. The impact to the CA of changes arising from updated performance assessment (PA) baselines [i.e., Saltstone Disposal Facility (SDF), E-Area Low-Level Waste Facility (ELLWF), and F & H Tank Farm (FTF & HTF) closures] is expected to be minor for the following reasons: The primary contributors to the SRS CA dose impact at the Upper Three Runs (UTR) point of assessment (POA) are the H-Canyon and Mixed Waste Management Facility (MWMF), contributing 68% and 9%, respectively, to the dose impact at that POA. The combined contribution to the UTR dose impact from all PA’s (SDF, ELLWF, FTF and HTF) is ~2% of this total. The 2010 SRS CA model validation performed indicates that the CA projected dose, while generally conservative, provides a reasonable representation of the maximum annual doses. Doses evaluated are well below the SRS established 15 mrem/yr administrative limit (Crapse et al. 2011). Based on the assessment presented within this annual review and collective engineering judgement, the conclusions of the 2010 SRS CA remain valid and there is reasonable assurance that SRS will meet the performance objectives delineated in DOE Manual 435.1-1. The 2010 SRS CA should be updated to incorporate PA changes, proposed changes to inventories and sources and model improvements accumulated since the 2010 CA. The timing will be dependent on the completion of the ongoing E-Area PA revision.

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

Transformative Technology for FLASH Radiation Therapy

The general concept of radiation therapy used in conventional cancer treatment is to increase the therapeutic index by creating a physical dose differential between tumors and normal tissues through precision dose targeting, image guidance, and radiation beams that deliver a radiation dose with high conformality, e.g., protons and ions. However, the treatment and cure are still limited by normal tissue radiation toxicity, with the corresponding side effects. A fundamentally different paradigm for increasing the therapeutic index of radiation therapy has emerged recently, supported by preclinical research, and based on the FLASH radiation effect. FLASH radiation therapy (FLASH-RT) is an ultra-high-dose-rate delivery of a therapeutic radiation dose within a fraction of a second. Experimental studies have shown that normal tissues seem to be universally spared at these high dose rates, whereas tumors are not. While dose delivery conditions to achieve a FLASH effect are not yet fully characterized, it is currently estimated that doses delivered in less than 200 ms produce normal-tissue-sparing effects, yet effectively kill tumor cells. Despite a great opportunity, there are many technical challenges for the accelerator community to create the required dose rates with novel compact accelerators to ensure the safe delivery of FLASH radiation beams.

43 PARTICLE ACCELERATORS↗
DOE OSTI2022

Transformative Technology for FLASH Radiation Therapy: A Snowmass 2021 White Paper

Conventional cancer therapies include surgery, radiation therapy, chemotherapy, and, more recently, immunotherapy. These modalities are often combined to improve the therapeutic index. The general concept of radiation therapy is to increase the therapeutic index by creating a physical dose differential between tumors and normal tissues through precision dose targeting, image guidance, and high radiation beams that deliver radiation dose with high conformality, e.g., protons and ions. However, treatment and cure are still limited by normal tissue radiation toxicity, with many patients experiencing acute and long-term side effects. Recently, however, a fundamentally different paradigm for increasing the therapeutic index of radiation therapy has emerged, supported by preclinical research, and based on the FLASH radiation effect. FLASH radiation therapy (FLASH-RT) is an ultra-high dose-rate delivery of a therapeutic radiation dose within a fraction of a second. Experimental studies have shown that normal tissues seem to be universally spared at these high dose rates, whereas tumors are not. The dose delivery conditions are not yet fully characterized. Still, it is currently estimated that large doses of 10 Gy or more delivered in 200 ms or less produce normal tissue sparing effects yet effectively kill tumor cells. There is a great opportunity, but also many technical challenges, for the accelerator community to create the required dose rates with novel and compact accelerators to ensure the safe delivery of FLASH radiation beams.

43 PARTICLE ACCELERATORS↗
DOE OSTI2023

Early-Stage Radiation Safety Analysis for the Spallation Neutron Source Second Target Station Bunker Operations

The Second Target Station project at Oak Ridge National Laboratory will develop a cold neutron source to meet growing experimental needs. This paper describes calculations of the residual dose rates associated with the monolith shield plug and the beamline bunker, two key conventional operations and radiation safety features. While neutron production is active, the instrument hall outside the bunker must be generally accessible with dose rates of less than 0.25 mrem/h. When neutron production is halted, the bunker must be accessible for hands-on maintenance operations. Further, these two requirements form the cause for the assessments reported herein of residual dose rates caused by the monolith shield plug and residual dose rates in the bunker. The monolith shield plug was shown to not produce significant dose rates inside the bunker after a 20-year lifetime, and the residual dose rates inside the bunker for the case of an operating beamline were shown to reasonably allow for hands-on maintenance. These calculations are based on preliminary design models of the relevant systems. Additionally, an example showing the significance of considering neutron supermirror physics in transport calculations that track nuclide production and destruction rates to produce gamma sources for residual dose rate calculations is included. The example shows that if neutron supermirror physics is not considered, dose rate fields may be significantly underpredicted.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2021

Simulation and Modeling of Time-Resolved X-Ray Detector for the Saturn Accelerator

In this work, we present the technology-aided computer design (TCAD) device simulation and modeling of a silicon p-i-n diode for detecting time-dependent X-ray radiation. We show that the simulated forward and reverse breakdown current–voltage characteristics agree well with the measured data under nonradiation environment by only calibrating carrier lifetimes for the forward bias case and avalanche model critical fields for the reverse bias condition. Using the calibrated parameters and other nominal material properties, we simulated the radiation responses of the p-i-n diode and compared with experimental data when the diode was exposed to X-ray radiation at Sandia’s Saturn facility and the Idaho State University (ISU) TriMeV facility. For Saturn’s Gaussian dose-rate pulses, we show three findings from TCAD simulations. First, the simulated photocurrents are in excellent agreement with the measured data for two dose-rate pulses with peak values of 1.16×10 10 and 1.88×10 10 rad(Si)/s. Second, the simulation results of high dose-rate pulses predict increased delayed photocurrents with longer time tails in the diode electrical responses due to excess carrier generation. Third, simulated peak values of diode radiation responses versus peak dose rates at different bias conditions provide useful guidance to determine the dose-rate range that the p-i-n diode can reliably detect in experiment. For TriMeV’s non-Gaussian dose-rate pulse, our simulated diode response is in decent agreement with the measured data without further calibration. We also studied the effects of device geometry, recombination process, and dose-rate enhancement via TCAD simulations to understand the higher measured response in the time after the peak dose-rate radiation for the p-i-n diode exposed to TriMeV irradiation.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2020

Update to the Performance Assessment for the Savannah River Site Saltstone Disposal Facility - 20124

In 2019, Savannah River Remediation developed a revision to the performance assessment (PA) on behalf of the U.S. Department of Energy (DOE) Savannah River Operations Office (SR) for the near-surface disposal of low-level waste at the Savannah River Site (SRS) Saltstone Disposal Facility (SDF). Soluble waste from SRS Tank Farms undergoes salt processing to remove cesium and other high activity constituents. The low-activity decontaminated salt solution (DSS) is then immobilized by mixing it into a cementitious waste form known as saltstone. After mixing, the saltstone is poured into leak-tight concrete vaults, known as saltstone disposal units (SDUs), where the waste form cures. By the time of facility closure, the SDF is expected to consist of 15 SDUs with a combined capacity of 1.06 E+09 L (280 Mgal) of cured saltstone. The facility operates under a Disposal Authorization Statement (DAS) from DOE and a permit from the South Carolina Department of Health and Environmental Control (SCDHEC). Since the start of operations in 1990, the SDF has received almost 6.7 E+07 L (18 Mgal) of DSS, resulting in the safe disposal of 2.7 E+16 Bq (7.3 E+05 Ci) of activity. Due to the radioactive decay of short-lived contaminants, the total remaining activity in the disposed waste is estimated to be approximately 1.4 E+16 Bq (3.9 E+05 Ci), as of September 2018. The Disposal Authorization Statement requires a demonstration that the system of engineered and natural features of the disposal facility will limit releases from the facility and be protective of human health and the environment for at least the next 1,000 years. The long-term performance of the facility was evaluated under the requirements of the DoE's Radioactive Waste Management Manual (US DOE Manual 435.1-1). Simulations were performed to demonstrate that the disposal facility would meet performance objectives specified in the manual. The evaluation was based on numerical models that simulate the releases of contaminants from the saltstone waste form. Contaminants were transported through groundwater and air pathways to points of assessment to evaluate compliance. In addition, the potential consequences of an inadvertent human intrusion (IHI) were also evaluated. In accordance with the guidance and recommendations in US DoE's technical standard for DAS, deterministic and probabilistic analyses were performed to demonstrate the SDF system, which includes the engineered cover system, the SDUs, the waste form, and the natural features of the site, provides a reasonable expectation that saltstone disposal will meet performance objectives. Post-closure doses to future members of the public (MOP) were evaluated; dose estimates from the air pathway were well below the 1.0 E-04 Sv/yr (10 mrem/yr) performance objective within 1,000 years, and the calculated doses to the MOP from all exposure pathways, including the groundwater and air pathways, were well below the 2.5 E-04 Sv/yr (25 mrem/yr) performance objective within 1,000 years. Doses following an assumed intrusion event within the facility boundaries were well below the IHI performance objectives, where the acute IHI dose was below the acute IHI dose performance objective of 5.0 E-03 Sv (500 mrem) and the chronic IHI dose was below the chronic IHI dose performance objective of 1.0 E-03 Sv/yr (100 mrem/yr) within 1,000 years. (author)

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

Radiation shielding analysis of the concrete hutch of a long beamline of the Advanced Photon Source Upgrade project

The radiation shielding analysis of the external concrete enclosure of a feature long beamline of the Advanced Photon Source Upgrade project is described in this work. The beamline has 2 undulators that can operate in two different configurations and the source terms are estimated for the worst configuration using the SRW code in the OASYS package as well as the STAC8 code. The shielding analysis is carried out using the STAC8 code, and for the case of pink beam, with the FLUKA code also. Effective dose rates on contact with the outside surface of the concrete walls for pink, reflected full spectrum and monochromatic beams are calculated. For the monochromatic beams, 2 sets of discrete bandwidths (BW) from the Double Crystal Monochromator (DCM) and the full reflectance of the Double Multilayer Monochromator (DMM) are used. Further, due to the size of the direct beam and the energy domain of interest in the direct and scattered fields, the importance of the equivalent dose to the lens of the eye is discussed with respect to the shielding calculations of such facilities. The dose rates outside the concrete hutch in the lateral directions calculated using STAC8 is about 50% higher than the FLUKA results for the case of pink beam but may be considered as a reasonable agreement due to the differences in the calculational methodologies. Those dose rates in the lateral and forward directions are higher for the pink beam to allow full occupancy from the radiation protection point of view. For the full spectrum after 3 reflections, the effective dose rates are less than 0.5 Sv/h in the lateral direction but is higher in the forward region below about 22° with respect to the beam direction. For the DCM beams, except for the scenario with one mirror and 0.1%BW, all dose rates are less than 0.5 Sv/h outside the concrete hutch in the lateral directions. Here, the higher harmonics are what contributes to the dose outside the shield and the bandwidths assumed to derive the photon flux thus becomes important. The dose rates with the DMM are generally lower compared to the 0.1% BW DCM beam but are higher than the results obtained with XOP calculated bandwidths.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

Phase IB study of ziv-aflibercept plus pembrolizumab in patients with advanced solid tumors

Background The combination of antiangiogenic agents with immune checkpoint inhibitors could potentially overcome immune suppression driven by tumor angiogenesis. We report results from a phase IB study of ziv-aflibercept plus pembrolizumab in patients with advanced solid tumors. Methods This is a multicenter phase IB dose-escalation study of the combination of ziv-aflibercept (at 2–4 mg/kg) plus pembrolizumab (at 2 mg/kg) administered intravenously every 2 weeks with expansion cohorts in programmed cell death protein 1 (PD-1)/programmed death-ligand 1(PD-L1)-naïve melanoma, renal cell carcinoma (RCC), microsatellite stable colorectal cancer (CRC), and ovarian cancer. The primary objective was to determine maximum tolerated dose (MTD) and recommended dose of the combination. Secondary endpoints included overall response rate (ORR) and overall survival (OS). Exploratory objectives included correlation of clinical efficacy with tumor and peripheral immune population densities. Results Overall, 33 patients were enrolled during dose escalation (n=3) and dose expansion (n=30). No dose-limiting toxicities were reported in the initial dose level. Ziv-aflibercept 4 mg/kg plus pembrolizumab 2 mg/kg every 2 weeks was established as the MTD. Grade ≥3 adverse events occurred in 19/33 patients (58%), the most common being hypertension (36%) and proteinuria (18%). ORR in the dose-expansion cohort was 16.7% (5/30, 90% CI 7% to 32%). Complete responses occurred in melanoma (n=2); partial responses occurred in RCC (n=1), mesothelioma (n=1), and melanoma (n=1). Median OS was as follows: melanoma, not reached (NR); RCC, 15.7 months (90% CI 2.5 to 15.7); CRC, 3.3 months (90% CI 0.6 to 3.4); ovarian, 12.5 months (90% CI 3.8 to 13.6); other solid tumors, NR. Activated tumor-infiltrating CD8 T cells at baseline (CD8+PD1+), high CD40L expression, and increased peripheral memory CD8 T cells correlated with clinical response. Conclusion The combination of ziv-aflibercept and pembrolizumab demonstrated an acceptable safety profile with antitumor activity in solid tumors. The combination is currently being studied in sarcoma and anti-PD-1-resistant melanoma. Trial registration number NCT02298959 .

Rahma, Osama E.↗
DOE OSTI2020

PNNL Measurement Results for the 2018 FlatTop Criticality Nuclear Accident Dosimetry Exercise at the Nevada National Security Site (IER-253)

The Pacific Northwest National Laboratory (PNNL) participated in the FlatTop exercise IER-253 at the National Nuclear Security Site (NNSS) during the week of May 21, 2018. This report summarizes the measurements of the personal nuclear accident dosimeters (PNAD). Additionally, portable survey instrument readings and measurements of the simulated biological samples are included. The FlatTop exercise (IER-253) tested only PNADs and did not include fixed nuclear accident dosimeters (FNAD). Previous PNAD results obtained in 2016 at the Godiva-IV exercise IER-148 were summarized in PNNL Measurement Results for the 2016 Criticality Accident Dosimetry Exercise at the Nevada National Security Site IER-148 (PNNL-26497). The results clearly indicated that the previous dose conversions were a historical legacy of the Hanford Site and corresponded closer to kerma dose (or first collision dose, and not to Hp(10) dose). The technical basis for PNAD dose was based on the Godiva-IV exercise and documented in A New Dose Calculation Methodology for New PNAD and FNAD Designs at PNNL (PNNL-27023). The configuration of the PNAD was described in these two documents. The FlatTop exercise was the first time PNNL tested this new dose calculation methodology. Results were generally favorable with some results conservatively over reported that fell above the performance criteria. The reported gamma dose results fell non-conservatively low because only the default parameters obtained from the Godiva-IV exercise were used, and because the OSL gamma dosimeter has an energy dependency. The results from the simulated biological samples were also favorable, and in several cases conservatively over reported.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2024

Modeling Plutonium Decorporation in a Female Nuclear Worker Treated with Ca-DTPA after Inhalation Intake

The present work models plutonium (Pu) biokinetics in a female former nuclear worker. Her bioassay measurements are available at the US Transuranium and Uranium Registries. The worker was internally exposed to a plutonium-americium mixture via acute inhalation at a nuclear weapons facility. She was medically treated with injections of 1 g Ca-DTPA on days 0, 5, and 14 after the intake. Between days 0 and 20, fecal and urine samples were collected and analyzed for 239 Pu and 241 Am. Subsequently, she was followed up for bioassay monitoring over 14 y, with additional post-treatment urine samples collected and analyzed for 239 Pu. The uniqueness of this dataset is due to the availability of: (1) both early and long-term bioassay data from a female with plutonium intake; (2) data on chelation therapy for a female; and (3) fecal measurement results. Chelation therapy with Ca- and/or Zn-salts of DTPA is known to aid in reducing the internal radiation dose by enhancing the excretion of plutonium and americium from the body. Such enhancement affects plutonium biokinetics in the human body, posing a challenge to the internal dose assessment. The current radiation dose assessment practice is to exclude the data affected by Ca-DTPA from the analysis. The present analysis is the first to explicitly model the chelation-affected bioassay data in a female by using a newly developed chelation model. Thus, the bioassay data collected during and after the Ca-DTPA administrations were used for biokinetic modeling and dose assessment. The Markov Chain Monte Carlo method was used to investigate model parameter uncertainty, based on the bioassay data and assumed prior probability distributions. A χ 2 /nData (number of data points) ≈ 1 was observed in this study, which indicates self-consistency of the data with the model. Results of this study show that the worker’s 239 Pu intake was 12 Bq, with a committed effective dose to the whole-body of 1.2 mSv and a committed equivalent dose to the bone surfaces, liver, and lungs of 37.8, 9.1, and 0.8 mSv, respectively. This study also discusses the worker’s dose reduction due to chelation treatment.

61 RADIATION PROTECTION AND DOSIMETRY↗