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All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

Formerly Used Sites Remedial Action Program (FUSRAP) W. R. Grace Building 23 Remedial Action-Challenges and Successes - 12247

Monazite sand processing was conducted at the W. R. Grace Curtis Bay Facility (Baltimore, Maryland) from mid-May 1956 through the spring of 1957 under license to the Atomic Energy Commission (AEC), for the extraction of source material in the form of thorium, as well as rare earth elements. The processing was conducted in the southwest quadrant of a ca. 100 year old, five-story, building (Building 23) in the active manufacturing portion of the facility. Building components and equipment in the southwest quadrant of Building 23 exhibited residual radiological activity remaining from the monazite sand processing. U.S. Army Corps of Engineers (USACE) conducted a remedial investigation (RI) and feasibility study (FS) and prepared a Record of Decision (ROD) to address residual radioactivity on building components and equipment in the southwest quadrant of Building 23. The remedy selected for the southwest quadrant of Building 23, which was documented in the ROD (dated May 2005), was identified as 'Alternative 2: Decontamination With Removal to Industrial Use Levels'. The selected remedy provided for either decontaminating or removing areas of radioactivity to meet the RGs. Demonstration of compliance with the selected ARAR was performed using the Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM) and other appropriate guidance, as well as appropriate dose modeling codes where necessary. USACE-Baltimore District along with its private industry partner worked together under the terms of a 2008 Settlement Agreement to implement the remedial action (RA) for the southwest quadrant of Building 23. The RA was conducted in two phases: Phase 1 was completed to improve the building condition for support of subsequent remedial action and decrease scope uncertainty of the remedial action, and Phase 2 included decontamination and removal activities to meet the RGs and demonstration of compliance with the selected ARAR. Challenges encountered during the RA include: coordination with stakeholders, coordination between multiple RA contractors, addressing unique structural challenges for Building 23, nonradiological hazards associated with the RA, weather issues, and complex final status survey (FSS) coordination. The challenges during the Phase 1 RA were handled successfully. The challenges for the Phase 2 RA, which is anticipated to be complete by late-summer of 2012, have been handled successfully so far. By fall of 2012, USACE is expecting to finalize a robust RA Closure Report, including the Final Status Survey Report, which summarizes the RA activities and documents compliance with the ROD. During the ongoing RA at Building 23, there have been and still are many challenges both technically and from a project management perspective, due in part to the nature and extent of impact at the site (residual radioactivity within an active processing building), dual oversight by the property owner and USACE, and site-specific challenges associated with a complex RA and multiple contractors. Currently, USACE and its industry partner are overseeing the completion of RA field activities. RA closure documentation for the remediation of Building 23 to address residual contamination in building materials will be reviewed/approved by USACE and its industry partner upon completion of the field activities. USACE and its industry partner are working well together, through the Settlement Agreement, to conduct a cost-efficient and effective remedial action to address the legacy issues at Building 23. This cooperative effort has set a firm foundation for achieving a successful RA at the RWDA using a 'forward think' approach, and it is a case study for other sites where an industry partner is involved. The collaborative effort led to implementation of an RA which is acceptable to the site owner, the regulators, and the public, thus allowing USACE to move this project forward successfully in the FUSRAP program. (authors)

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

TU-D-209-05: Automatic Calculation of Organ and Effective Dose for CBCT and Interventional Fluoroscopic Procedures

Purpose: To compare PCXMC and EGSnrc calculated organ and effective radiation doses from cone-beam computed tomography (CBCT) and interventional fluoroscopically-guided procedures using automatic exposure-event grouping. Methods: For CBCT, we used PCXMC20Rotation.exe to automatically calculate the doses and compared the results to those calculated using EGSnrc with the Zubal patient phantom. For interventional procedures, we use the dose tracking system (DTS) which we previously developed to produce a log file of all geometry and exposure parameters for every x-ray pulse during a procedure, and the data in the log file is input into PCXMC and EGSnrc for dose calculation. A MATLAB program reads data from the log files and groups similar exposures to reduce calculation time. The definition files are then automatically generated in the format used by PCXMC and EGSnrc. Processing is done at the end of the procedure after all exposures are completed. Results: For the Toshiba Infinix CBCT LCI-Middle-Abdominal protocol, most organ doses calculated with PCXMC20Rotation closely matched those calculated with EGSnrc. The effective doses were 33.77 mSv with PCXMC20Rotation and 32.46 mSv with EGSnrc. For a simulated interventional cardiac procedure, similar close agreement in organ dose was obtained between the two codes; the effective doses were 12.02 mSv with PCXMC and 11.35 mSv with EGSnrc. The calculations can be completed on a PC without manual intervention in less than 15 minutes with PCXMC and in about 10 hours with EGSnrc, depending on the level of data grouping and accuracy desired. Conclusion: Effective dose and most organ doses in CBCT and interventional radiology calculated by PCXMC closely match those calculated by EGSnrc. Data grouping, which can be done automatically, makes the calculation time with PCXMC on a standard PC acceptable. This capability expands the dose information that can be provided by the DTS. Partial support from NIH Grant R01-EB002873 and Toshiba Medical Systems Corp.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2018

Accident Scenarios for IOTA Ring and Dose Distributions Calculated with MARS15 Code

The Integrable Optics Test Accelerator (IOTA) at Fermilab’s Accelerator Science and Technology (FAST) facility [1] is a storage ring for advanced high intensity beam physics research. Layout of the ring is shown in Fig. 1. A plan view of the ring with enclosure walls are shown in Fig. 2. The entire beam, stored in the IOTA ring, is assumed to be completely lost at either of the three locations shown in Fig. 2. Therefore, we will consider three separate corresponding accident scenarios. The maximal amount of electrons stored in the IOTA ring is estimated to be 2×10 10 electrons with energy of 150 MeV. At locations 1 and 2 the beam is lost on 60-deg dipole walls, while at location 3 the beam is lost on Lambertson magnet wall.

43 PARTICLE ACCELERATORS↗
DOE OSTI2023

Shutdown dose rate analysis with the Shift Monte Carlo radiation transport code and modular verification workflow

Calculation of the shutdown dose rate is crucial for safe fusion reactor operations. The Rigorous-two-step (R2S) method is a method that requires connected neutron transport, activation, and gamma transport. Shift has integrated variance reduction with a deterministic solver Denovo, supports multiple geometry formats, and is scalable. These features make it an attractive transport solver choice for an R2S workflow. An R2S workflow for the Shift Monte Carlo code is developed and compared to the existing Oak Ridge National Laboratory Shutdown Dose Rate Code Suite (ORCS) workflow. Also, a Python framework for integrating two R2S workflows is developed to mix and match each step in the R2S workflow for improved collaboration and verification experience. In this study, results show that the Shift-Denovo R2S workflow and the ORCS workflow calculate the shutdown dose rate of the ITER Shutdown Dose Rate benchmark problem with an average relative error of 2.285%.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2014

Use of the portable decay gamma source feature in the R2Smesh code system for the accurate estimation of photon radiation dose fields around activated ITER components - Paper 111

The R2Smesh ('Rigorous 2-step') code system, developed by KIT for the determination of shut-down dose rate distribution on mesh grid, can be utilized to provide photon radiation dose field around activated components which are moved from the irradiation site in fusion reactor to another external location. R2Smesh couples MCNP transport calculations and FISPACT inventory calculations through suitable interfaces utilizing MCNP mesh tally capability. Neutron flux and the decay gamma source are provided with high resolution on the mesh grid superimposed to the real geometry to take into account the spatial variations of the flux and the decay gamma source distributions. This feature enables the decay gamma source distribution to be transferred from the irradiation site in the reactor to any other external location for the determination of photon flux and dose rate distributions. In this work, we applied the feature of the portable decay gamma source in R2Smesh code system to a dose rate analysis for a tunnel connecting the ITER tokamak complex building (TCB) and the hot cell facility building (HCFB). This paper presents the use of the portability feature of the decay gamma source in R2Smesh interface code system developed by KIT for a dose rate analysis of the ITER tunnel connecting TCB and HCFB. This application shows that the decay gamma source of the activated components could be transferred successfully from one coordinate system to another coordinate system and superimposed to the geometry of the components. This is very useful when the activated components need to be transported from the reactor to somewhere else for maintenance, testing or storage and the dose rate analysis is requested. Another portable feature, which could convert the 3D decay gamma source distribution generated by R2Smesh system to the MCNP 'SDEF' card source description and avoid the use of source sampling subroutine, is under consideration. With this feature, one decay gamma source result for the specific reactor components could be shared by different institutions who need to use the same decay gamma source for different analysis. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2023

Findings during Incorporation of the Federal Guidance Report No. 15 External Dose Coefficients into the RESRAD Family of Codes

Argonne National Laboratory was tasked by the U.S. Department of Energy (DOE) to incorporate the recently published Federal Guidance Report No. 15 (FGR 15) (EPA 2019) external dose rate coefficients into the database of the RESRAD family of codes. FGR 15 is one of two publications that contain age-specific external effective dose rate coefficients that are based on the International Commission on Radiological Protection (ICRP) Publication 103 (ICRP-103) methodology (ICRP 2007) and the ICRP-107 (ICRP 2008) nuclear decay data; ICRP-144 (ICRP 2020) is the other publication.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2014

Final Radiological Assessment of External Exposure for CLEAR-Line Americium Recovery Operations

Los Alamos National Laboratory is currently planning to implement an americium recovery program. The americium, ordinarily isotopically pure 241Am, would be extracted from existing Pu materials, converted to an oxide and shipped to support fabrication of americium oxide-beryllium neutron sources. These operations would occur in the currently proposed Chloride Extraction and Actinide Recovery (CLEAR) line of glove boxes. This glove box line would be collocated with the currently-operational Experimental Chloride Extraction Line (EXCEL). The focus of this document is to provide an in-depth assessment of the currently planned radiation protection measures and to determine whether or not further design work is required to satisfy design-goal and ALARA requirements. Further, this document presents a history of americium recovery operations in the Department of Energy and high-level descriptions of the CLEAR line operations to provide a basis of comparison. Under the working assumptions adopted by this study, it was found that the evaluated design appears to mitigate doses to a level that satisfies the ALARA-in-design requirements of 10 CFR 835 as implemented by the Los Alamos National Laboratory procedure P121. The analyses indicate that extremity doses would also meet design requirements. Dose-rate calculations were performed using the radiation transport code MCNP5 and doses were estimated using a time-motion study developed in consort with the subject matter expert. A copy of this report and all supporting documentation are located on the Radiological Engineering server at Y:\Rad Engineering\2013 PROJECTS\TA-55 Clear Line.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2024

PUFFIn – A user friendly fast interface for calculating and visualizing the dose distribution in materials

This article describes a new graphical user interface that uses the PENELOPE Monte Carlo code to calculate dose distributions in materials. It is named PUFFIn, for the Penelope User-Friendly Fast Interface and was developed as an educational and scoping analysis tool for non-experts in radiation modeling. PUFFIn enables the user to visualize and compare the dose distributions in objects (e.g., sterilized healthcare products) irradiated with cobalt-60 gamma-rays, electron beam (E-beam) or X-rays. From such comparisons, the user can determine the most efficient product and/or packaging designs for any given radiation field – whether for conceptual or existing products and packaging. PUFFIn is distributed in a complete, self-contained package of software, including the PENELOPE radiation transport code, a standard graphics package, and a set of simple exercises. The package is available to any user at no cost and requires a minimal amount of training compared to other similar software. PUFFIn's capabilities are described, as well as validation measurements performed at Texas A&M University E-beam facility and the Aerial E-beam Facility in France.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2010

Effects of breast-air and breast-lung interfaces on the dose rate at the planning target volume of a MammoSite catheter for Yb-169 and Ir-192 HDR sources

Purpose: To study the effects of the breast-air and breast-lung interfaces on the absorbed dose within the planning target volume (PTV) of a MammoSite balloon dose delivery system as well as the effect of contrast material on the dose rate in the PTV. Methods: The Monte Carlo MCNP5 code was used to simulate dose rate in the PTV of a 2 cm radius MammoSite balloon dose delivery system. The simulations were carried out using an average female chest phantom (AFCP) and a semi-infinite water phantom for both Yb-169 and Ir-192 high dose rate sources for brachytherapy application. Gastrografin was introduced at varying concentrations to study the effect of contrast material on the dose rate in the PTV. Results: The effect of the density of the materials surrounding the MammoSite balloon containing 0% contrast material on the calculated dose rate at different radial distances in the PTV was demonstrated. Within the PTV, the ratio of the calculated dose rate for the AFCP and the semi-infinite water phantom for the point closest to the breast-air interface (90 deg.) is less than that for the point closest to the breast-lung interface (270 deg.) by 11.4% and 4% for the HDR sources of Yb-169 and Ir-192, respectively. When contrast material was introduced into the 2 cm radius MammoSite balloon at varying concentrations, (5%, 10%, 15%, and 20%), the dose rate in the AFCP at 3.0 cm radial distance at 90 deg. was decreased by as much as 14.8% and 6.2% for Yb-169 and Ir-192, respectively, when compared to that of the semi-infinite water phantom with contrast concentrations of 5%, 10%, 15%, and 20%, respectively. Conclusions: Commercially available software used to calculate dose rate in the PTV of a MammoSite balloon needs to account for patient anatomy and density of surrounding materials in the dosimetry analyses in order to avoid patient underdose.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2016

SU-F-T-364: Monte Carlo-Dose Verification of Volumetric Modulated Arc Therapy Plans Using AAPM TG-119 Test Patterns

Purpose: To investigate the Monte Carlo (MC)-based dose verification for VMAT plans by a treatment planning system (TPS). Methods: The AAPM TG-119 test structure set was used for VMAT plans by the Pinnacle3 (convolution/superposition), using a Synergy radiation head of a 6 MV beam with the Agility MLC. The Synergy was simulated with the EGSnrc/BEAMnrc code, and VMAT dose distributions were calculated with the EGSnrc/DOSXYZnrc code by the same irradiation conditions as TPS. VMAT dose distributions of TPS and MC were compared with those of EBT3 film, by 2-D gamma analysis of ±3%/3 mm criteria with a threshold of 30% of prescribed doses. VMAT dose distributions between TPS and MC were also compared by DVHs and 3-D gamma analysis of ±3%/3 mm criteria with a threshold of 10%, and 3-D passing rates for PTVs and OARs were analyzed. Results: TPS dose distributions differed from those of film, especially for Head & neck. The dose difference between TPS and film results from calculation accuracy for complex motion of MLCs like tongue and groove effect. In contrast, MC dose distributions were in good agreement with those of film. This is because MC can model fully the MLC configuration and accurately reproduce the MLC motion between control points in VMAT plans. D95 of PTV for Prostate, Head & neck, C-shaped, and Multi Target was 97.2%, 98.1%, 101.6%, and 99.7% for TPS and 95.7%, 96.0%, 100.6%, and 99.1% for MC, respectively. Similarly, 3-D gamma passing rates of each PTV for TPS vs. MC were 100%, 89.5%, 99.7%, and 100%, respectively. 3-D passing rates of TPS reduced for complex VMAT fields like Head & neck because MLCs are not modeled completely for TPS. Conclusion: MC-calculated VMAT dose distributions is useful for the 3-D dose verification of VMAT plans by TPS.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2014

SU-E-T-506: Dosimetric Verification of Photon MLC Delivered Electron Fields for Implementing MERT On An Artiste Linac

Purpose: To verify the dose accuracy of photon MLC delivered electron fields for implementing energy-intensity modulated electron radiotherapy (MERT) on an Artiste linac. Methods: It was proposed to deliver MERT on an Artiste linac at a short SSD (60 cm) to reduce beam penumbra caused by electron scatters. An in-house developed Monte Carlo (MC)-based dose calculation/optimization planning code was used for treatment planning. Our previous study showed that the measured dose distribution of a breast plan showed good agreement with the calculations in low-medium dose regions while the differences in high dose regions were outstanding. A continuous work found that the discrepancy is mainly caused by improper modeling in MC for the single focused MLC in the Artiste which was simplified as double focused in the previous MC simulations. With this remodeled MLC in the calculations, an energy-intensity modulated electron plan using 6, 9, 12 and 15 MeV was generated for a breast treatment on a breast phantom at a 60 cm SSD and recalculated regarding a solid water phantom. For a test study, four of MLC segments (each with a different energy) generated in the plan were delivered to the phantom and a film measurement was performed at the depth of 2 cm. The measured 2D dose distribution was then compared with calculations. Results: For composite doses of the four segments, measured 2D dose distributions overall agree well with the calculations (3mm/3%) in most area. The separate measurement for a single MLC segment for each of energies also showed the consistence with the calculations. Conclusion: After remodeling MLC in the MC calculations, the measured dose distribution for a subset of MLC segments from a MERT plan showed good agreement. Further detailed verification for the full plan will be the work in the next step.

07 ISOTOPE AND RADIATION SOURCES↗
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 OSTI2015

WE-AB-204-11: Development of a Nuclear Medicine Dosimetry Module for the GPU-Based Monte Carlo Code ARCHER

Purpose: To develop a nuclear medicine dosimetry module for the GPU-based Monte Carlo code ARCHER. Methods: We have developed a nuclear medicine dosimetry module for the fast Monte Carlo code ARCHER. The coupled electron-photon Monte Carlo transport kernel included in ARCHER is built upon the Dose Planning Method code (DPM). The developed module manages the radioactive decay simulation by consecutively tracking several types of radiation on a per disintegration basis using the statistical sampling method. Optimization techniques such as persistent threads and prefetching are studied and implemented. The developed module is verified against the VIDA code, which is based on Geant4 toolkit and has previously been verified against OLINDA/EXM. A voxelized geometry is used in the preliminary test: a sphere made of ICRP soft tissue is surrounded by a box filled with water. Uniform activity distribution of I-131 is assumed in the sphere. Results: The self-absorption dose factors (mGy/MBqs) of the sphere with varying diameters are calculated by ARCHER and VIDA respectively. ARCHER’s result is in agreement with VIDA’s that are obtained from a previous publication. VIDA takes hours of CPU time to finish the computation, while it takes ARCHER 4.31 seconds for the 12.4-cm uniform activity sphere case. For a fairer CPU-GPU comparison, more effort will be made to eliminate the algorithmic differences. Conclusion: The coupled electron-photon Monte Carlo code ARCHER has been extended to radioactive decay simulation for nuclear medicine dosimetry. The developed code exhibits good performance in our preliminary test. The GPU-based Monte Carlo code is developed with grant support from the National Institute of Biomedical Imaging and Bioengineering through an R01 grant (R01EB015478)

60 APPLIED LIFE SCIENCES↗
DOE OSTI2016

TU-H-BRC-03: Evaluation of Very High-Energy Electron (VHEE) Beams in Comparison to VMAT and PBS Treatment Plans

Purpose: The aim of this study was to evaluate the performance of very high-energy electron (VHEE) beams in comparison to clinically delivered treatment plans generated with volumetric modulated arc therapy (VMAT) and proton pencil beam scanning (PBS) technology. Methods: Three clinical cases were selected (prostate, lung, and pediatric CNS). The VHEE plans were calculated in the Monte Carlo EGSnrc code and pencil beam doses were calculated using the DOSxyznrc MC code for 100 and 200 MeV beams. Treatment plans with VHEE, VMAT, and PBS were optimized in a research version of RayStation using an in house build script in order to minimize operator bias between the different techniques. Results: For the prostate cancer case, the PBS plan showed lower mean organ at risk (OAR) doses compared to the other modalities. An exception was the femoral heads, due to the lateral beam arrangements. The VMAT plan showed lower mean doses to the rectum and the bladder compared to the 100 MeV VHEE plan. The lung cancer case showed minor differences between the three modalities. However, the PBS plan showed a lower contralateral lung dose. The pediatric CNS case showed a better conformity and lower spinal cord dose for the 100 MeV VHEE plan. For all cases, the 200 MeV VHEE plans were found to be similar to or better than the 100 MeV VHEE plans. Conclusion: The present study showed that VHEE plans are similar or superior to VMAT plans with reduced mean OAR dose and increased target conformity for a variety of clinical cases. With increased VHEE energy, better conformity and even higher reductions in mean OAR doses can be achieved. Funding: DoD, Award#:W81XWH-13-1-0165, Weston Havens Foundation, Bio-X (Stanford University), the Office of the Dean of the Medical School, the Office of the Provost (Stanford University), and the Swedish Childhood Cancer Foundation. BL and PM are founders of TibaRay,Inc. BL and PM have received research grants from Varian and RaySearch Laboratory.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2018

Nonhuman biota dose rate estimation from liquid effluent releases during normal operations of light water reactors using the LADTAP II computer code

Here, the LADTAP II computer code was used to estimate the dose rates to seven nonhuman biota (fish, algae, invertebrate, muskrat, raccoon, heron, and duck) from liquid effluent releases during the normal operations of new light water reactor facilities in the United States. The estimated dose rates to nonhuman biota depended on the nuclear power plant design but were orders of magnitude less than the ‘derived consideration reference levels’ (DCRL) in International Commission on Radiological Protection (ICRP) Publication 108. The estimated dose rates were at least three orders of magnitude lower than the guideline values identified by the International Atomic Energy Agency (IAEA) and the National Council on Radiation Protection and Measurements to cause observable changes in populations of biota. Radionuclides that contributed most of the dose rates for seven nonhuman biota depended on the nuclear power plant design. The differences in the estimated dose rates to seven nonhuman biota calculated from the default bioaccumulation factors used in the LADTAP II computer code and the dose rates calculated from the concentration ratios of radionuclides in the IAEA Technical Report Series (TRS) wildlife handbook were less than one order of magnitude for most default biota in LADTAP. The concentration ratios for some radionuclides in liquid effluent releases from nuclear power plant designs that contributed more than 5% of the internal dose rates for invertebrates (P, and Ru) and algae (La and Pr) were not available in the IAEA TRS wildlife handbook.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2018

An Evidence-Based Critique of U.S. Nuclear Waste Policy Using Data from Reactor and Cyclotron Sites - 18568

Concrete at nuclear facilities becomes radioactive when it is exposed to a neutron flux, and when such facilities are vacated, decommissioning activities are enacted to make the site safe for reuse. These activities generate wastes that the public expects to be managed in their best interest. This paper provides side-by-side comparisons of concrete bio-shielding encountered at typical reactor and cyclotron decommissioning projects. Three sites are considered, and results from characterization samples are used to demonstrate that the wastes generated by such sites are similar in terms of content and risk. First, the dose potential represented by each waste is determined from mass-based radioactivity concentration with the RESRAD-BUILD computer code. The resulting doses are translated into population-risk estimates by applying dose-response predictions from the National Academy of Sciences' 2006 BEIR VII report. In terms of possible adverse health outcomes, the risk of solid cancer or leukemia associated with the wastes this paper examines is 8 to 50 extra cases per 100,000 persons. The regulatory framework applicable to each site is also reviewed, and the waste management strategies used at each site are contrasted. This paper demonstrates that the wastes it examines are virtually indistinguishable. Contrary to this finding, the wastes are regulated, managed, and disposed in very different manners according to current U.S. nuclear waste policy. The value of this paper is that the side-by-side comparisons it presents provides an opportunity to examine the risk-benefit of existing policies in terms of the benefits or disparities they represent. (author)

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

A Comprehensive Radiological Characterization of the ACRR Facility

Radiation Protection (628) in conjunction with Nuclear Facility Operations (1381) completed a comprehensive radiological conditions characterization of the ACRR facility between August 2015 and January 2017 to better understand the radiological environment in and around the ACRR High-Bay. The overall goals of the characterization, as identified by 1381 and 628, were to determine potential dose to workers during routine reactor operations, areas of potential elevated dose in and around the ACRR High-Bay, the relationship between reactor power and dose during steady-state operations, and the dose per pulse for pulsed operations. To accomplish this, eight configurations were identified of interest for characterization based on increased dose potential for which field surveys were completed to include assessment of the neutron spectra for each, determination of neutron dose conversion factors (Neutron Codes) based on defined neutron spectra, mapped dose surveys for each configuration, as well as determination of the relationship between dose and reactor power level during steady-state reactor operations. Steady-state survey data was also used to calculate dose per pulse approximations during pulsed reactor operations.

61 RADIATION PROTECTION AND DOSIMETRY↗