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18 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2014

Nuclear Data Sheets for A = 235

Spectroscopic data and level schemes from radioactive decay and nuclear reaction studies are presented here for all nuclei with mass number A=235. The highlight of this evaluation consists of the precise and comprehensive Coulomb excitation study (2012Wa35) on {sup 235}U, which in addition to the 7/2[743] ground state rotational band, extended the 1/2[631], 5/2[622], 5/2[752], and 3/2[631] rotational bands up to Jπ=53/2+, 49/2+, 441/2−, and 43/2+, respectively. This evaluation presents a study (2010Hu02) of the {sup 237}Np({sup 116}Sn, {sup 118}Snγ) reaction where the ground state rotational band 5/2[642] was observed up to Jπ=(53/2+). It is worth for historical knowledge to mention the report on the “Discovery of isotopes of the transuranium elements with 93≤Z≤98” (2013Fr02), where the information for elements Np, Pu, and Am with mass number A=235 is given. {sup 235}Cf has not been observed. The alpha hindrance factors (HF) presented in this evaluation were calculated using values of the radius parameter (r{sub 0}) interpolated from those for even–even adjacent nuclei given by 1998Ak04.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Options for the Removal of Cesium and Strontium from Acidic Uranyl Sulfate Solutions from Molybdenum-99 Production

A small laboratory-scale column process aimed at removing fission products from prototypical acidic sulfate solution wastes from a prototypical Mo-99 recovery process has been operated in a radioactive materials hood at Savannah River National Laboratory. This equipment has been operated with realistic concentrations of uranium (0.4 wt % 235 U), plutonium ( 239 Pu), neptunium ( 237 Np), non-radioactive surrogates of common fission products (including Cs and Sr) and two promising absorbents AMP-PAN and CST R91290-B. The goal of this work was to demonstrate the removal and concentration of high specific activity isotopes which are expected to dominate the classification of the low-level waste from the production of 99 Mo. Both AMP-PAN and CST R9120-B are attractive absorbents for the removal of 137 Cs from an acidic sulfate waste stream. They do not provide a solution for the removal of 90 Sr from the acidic solutions, but CST should be investigated at a higher pH. Limited removal of transuranic species and almost no removal of uranium was observed in this testing. This could be important due to the relatively large amount of uranium expected to be present in such waste streams. The capacity of these absorbents for 137 Cs will likely be dependent on the Cs concentration in the solutions processed but loading in the range of 0.2 to 5 mg Cs per gram of absorbent appears to be achievable.

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

Options for the Removal of Cesium and Strontium from Acidic Uranyl Sulfate Solutions from Molybdenum-99 Production: Part II The Effect of Sodium Carbonate Adjustment to pH 9 on Strontium Removal

A small laboratory-scale column process aimed at removing fission products from prototypical acidic sulfate solution wastes from a prototypical Mo-99 recovery process has been operated in a radioactive materials hood at Savannah River National Laboratory. This equipment has been operated with realistic concentrations of uranium (0.4 wt % 235 U), plutonium ( 239 Pu), neptunium ( 237 Np), non-radioactive surrogates of common fission products (including Cs and Sr) and two promising absorbents – AMP-PAN and CST R9120-B. The goal of this work was to demonstrate the removal and concentration of high specific activity isotopes which are expected to dominate the classification of the low-level waste from the production of 99 Mo. Both AMP-PAN and CST R9120-B are attractive absorbents for the removal of 137 Cs from an acidic sulfate waste stream. They do not provide a solution for the removal of 90 Sr from the acidic solutions. CST was shown to be useful at removing both Cs and Sr after adjustment of the acidic sulfate solution with sodium carbonate to pH 9. Limited removal of transuranic species and almost no removal of uranium was observed for CST R9120B absorbent. This could be important due to the relatively large amount of uranium expected to be present in such waste streams. The capacity of these absorbents for 137 Cs will likely be dependent on the Cs concentration in the solutions processed but loading in excess of 0.7 mg Cs per gram of CST absorbent while loading 0.14 mg Sr per gram of CST absorbent appears to be achievable from pH 9 carbonate solution.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

235 U– 231 Pa age dating of uranium materials for nuclear forensic investigations

Here, age dating of nuclear material can provide insight into source and suspected use in nuclear forensic investigations. We report here a method for the determination of the date of most recent chemical purification for uranium materials using the 235 U- 231 Pa chronometer. Protactinium is separated from uranium and neptunium matrices using anion exchange resin, followed by sorption of Pa to an SiO 2 medium. The concentration of 231 Pa is measured by isotope dilution mass spectrometry using 233 Pa spikes prepared from an aliquot of 237 Np and calibrated in-house using the rock standard Table Mountain Latite and the uranium isotopic standard U100. Combined uncertainties of age dates using this method are 1.5 to 3.5 %, an improvement over alpha spectrometry measurement methods. Model ages of five uranium standard reference materials are presented; all standards have concordant 235 U- 231 Pa and 234 U- 230 Th model ages.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2017

Impact of Reprocessed Uranium Management on the Homogeneous Recycling of Transuranics in PWRs

This article presents the results of a neutronics analysis related to the homogeneous recycling of transuranics (TRU) in PWRs with a MOX fuel using enriched uranium instead of depleted uranium. It also addresses an often, if not always, overlooked aspect related to the recycling of TRU in PWRs, namely the use of reprocessed uranium. From a neutronics point of view, it is possible to multi-recycle the entirety of the plutonium with or without neptunium and americium in a PWR fleet using MOX-EU fuel in between one third and two thirds of the fleet. Recycling neptunium and americium with plutonium significantly decreases the decay heat of the waste stream between 100 to 1,000 years compared to those of an open fuel cycle or when only plutonium is recycled. The uranium present in MOX-EU used fuel still contains a significant amount of 235 uranium and recycling it makes a major difference on the natural uranium needs. For example, a PWR fleet recycling its plutonium, neptunium and americium in MOXEU needs 28 percent more natural uranium than a reference UO 2 open cycle fleet generating the same energy if the reprocessed uranium is not recycled and 19 percent less if the reprocessed uranium is recycled back in the reactors, i.e. a 47 percent difference.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Short-lived fission product measurements from >0.1 MeV neutron-induced fission using boron carbide.

A boron carbide shield was designed, custom fabricated, and used to create a fast fission energy neutron spectrum. The fissionable isotopes 233, 235, 238U, 237Np, and 239Pu were separately placed inside of this shield and irradiated under pulsed conditions at the Washington State University 1 MW TRIGA reactor. A unique set of fission product gamma spectra were collected at short times (4 minutes to 1 week) post-fission. Gamma spectra were collected on single-crystal high purity germanium detectors and on Pacific Northwest National Laboratory's (PNNL's) Direct Simultaneous Measurement (DSM) system composed of HPGe detectors connected in coincidence. This work defines the experimental methods used to produce and collect the gamma data, and demonstrates the validity of the measurements. It is important to fully document this information so the data can be used with high confidence for the advancement of nuclear science and non-proliferation applications. The gamma spectra collected in these and other experiments will be made publicly available at https://spcollab.pnl.gov/sites/gammadata or via the link at http://rdnsgroup.pnl.gov. A revised version of this publication will be posted with the data to make the experimental details available to those using the data.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2011

Delayed Neutron and Delayed Photon Characteristics from Photofission of Actinides

Delayed neutron (DN) and delayed photon (DP) emissions from photofission reactions play an important role for applications involving nuclear material detection and characterization. To provide new, accurate, basic nuclear data for evaluations and data libraries, an experimental programme of DN and DP measurements has been undertaken for actinides with bremsstrahlung endpoint energy in the giant resonance region ({approx}15 MeV). In this paper, the experimental setup and the data analysis method will be described. Experimental results for DN and DP characteristics will be presented for {sup 232}Th, {sup 235,238}U, {sup 237}Np, and {sup 239}Pu. Finally, an example of an application to study the contents of nuclear waste packages will be briefly discussed.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2015

Characterization of Delayed-Particle Emission Signatures for Pyroprocessing. Part 1: ABTR Fuel Assembly

A three-part study is conducted using the MCNP6 Monte Carlo radiation-transport code to calculate delayed-neutron (DN) and delayed-gamma (DG) emission signatures for nondestructive assay (NDA) metal-fuel pyroprocessing. In Part 1, MCNP6 is used to produce irradiation-induced used nuclear fuel (UNF) isotopic inventories for an Argonne National Laboratory (ANL) Advanced Burner Test Reactor (ABTR) preconceptual design fuel assembly (FA) model. The initial fuel inventory consists of uranium mixed with light-water-reactor transuranic (TRU) waste and 10 wt% zirconium (U-LWR-SFTRU-10%Zr). To facilitate understanding, parametric evaluation is done using models for 3% and 5% initial 235 U a% enrichments, burnups of 5, 10, 15, 20, 30, …, 120 GWd/MTIHM, and 3-, 5-, 10-, 20-, and 30- year cooling times. Detailed delayed-particle radioisotope source terms for the irradiate FA are created using BAMF-DRT and SOURCES3A. Using simulation tallies, DG activity ratios (DGARs) are developed for 134 Cs/ 137 Cs 134 Cs/ 154 Eu, and 154 Eu/ 137 Cs markers as a function of (1) burnup and (2) actinide mass, including elemental uranium, neptunium, plutonium, americium, and curium. Spectral-integrated DN emission is also tallied. The study reveals a rich assortment of DGAR behavior as a function of DGAR type, enrichment, burnup, and cooling time. Similarly, DN emission plots show variation as a function of burnup and of actinide mass. Sensitivity of DGAR and DN signatures to initial 235 U enrichment, burnup, and cooling time is evident. Comparisons of the ABTR radiation signatures and radiation signatures previously reported for a generic Westinghouse oxide-fuel assembly indicate that there are pronounced differences in the ABTR and Westinghouse oxide-fuel DN and DG signatures. These differences are largely attributable to the initial TRU inventory in the ABTR fuel. The actinide and nonactinide inventories for the FA models serve as source materials for the pre- and postelectrorefining models to be reported in Parts 2 and 3.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2013

Cross sections and barriers for nuclear fission induced by high-energy nucleons

The cross sections for the fission of {sup 232}Th, {sup 235,238}U, {sup 237}Np, and {sup 239}Pu target nuclei that was induced by 20- to 1000-MeV neutrons and protons were calculated. The respective calculations were based on the multiconfiguration-fission (MCFx) model, which was used to describe three basic stages of the interaction of high-energy nucleons with nuclei: direct processes (intranuclear cascade), equilibration of the emerging compound system, and the decay of the compound nucleus (statistical model). Fission barriers were calculated within the microscopic approach for isotopic chains formed by 15 to 20 nuclei of the required elements. The calculated fission cross sections were compared with available experimental data. It was shown that the input data set and the theoretical model used made it possible to predict satisfactorily cross section for nuclear fission induced by 20- to 1000-MeV nucleons.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2014

Analysis on fuel breeding capability of FBR core region based on minor actinide recycling doping

Nuclear fuel breeding based on the capability of fuel conversion capability can be achieved by conversion ratio of some fertile materials into fissile materials during nuclear reaction processes such as main fissile materials of U-233, U-235, Pu-239 and Pu-241 and for fertile materials of Th-232, U-238, and Pu-240 as well as Pu-238. Minor actinide (MA) loading option which consists of neptunium, americium and curium will gives some additional contribution from converted MA into plutonium such as conversion Np-237 into Pu-238 and it's produced Pu-238 converts to Pu-239 via neutron capture. Increasing composition of Pu-238 can be used to produce fissile material of Pu-239 as additional contribution. Trans-uranium (TRU) fuel (Mixed fuel loading of MOX (U-Pu) and MA composition) and mixed oxide (MOX) fuel compositions are analyzed for comparative analysis in order to show the effect of MA to the plutonium productions in core in term of reactor criticality condition and fuel breeding capability. In the present study, neptunium (Np) nuclide is used as a representative of MAin trans-uranium (TRU) fuel composition as Np-MOX fuel type. It was loaded into the core region gives significant contribution to reduce the excess reactivity in comparing to mixed oxide (MOX) fuel and in the same time it contributes to increase nuclear fuel breeding capability of the reactor. Neptunium fuel loading scheme in FBR core region gives significant production of Pu-238 as fertile material to absorp neutrons for reducing excess reactivity and additional contribution for fuel breeding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

A Method for Understanding the Behaviour of Radioactive Decay Chains in Radiological Assessments - 19382

In many radiological groundwater pathway assessments, especially those where the assessment timescales of interest are relatively short (say a few thousand years or less), contributions to effective dose or radiological risk are dominated by radionuclides such as C-14, Cl-36 or I-129. Over longer timescales, additional radionuclides can also become key contributors to risk. In particular, members of the U-238 decay chain, notably Ra-226 and Pb-210, are likely to be key radionuclides. Other decay chains that can contribute significantly to dose or risk at times far into the future include Np-237 → U-233 → Th-229 and U-235 → Pa-231 → Ac-227. In this paper, a method is described for understanding decay chain behaviour in radiological assessments. The method uses a graphical approach in which the various modes of transport and decay in a system of connected compartments are represented as pathways between two nodes in a two-dimensional grid. Each pathway corresponds to a distinct transport and decay process. By so doing, the different transport and decay processes can be identified in a systematic way. The result is an algorithm for generating the contributions to effective dose or radiological risk without explicit solution of the transport equations. The method has been tested against numerical models and analytic solutions and found to give excellent agreement, while at the same time providing information about decay chain behaviour that would be difficult to obtain from a numerical assessment model. The power of the method is illustrated by applying it to some examples based on real repository concepts. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2010

Nuclear Resonance Fluorescence of 237 Np

Identification and characterization of fissile material is of international concern because of the possibility of proliferation. For safeguards applications, techniques providing isotopic information are necessary [1]. However, few techniques can distinguish between fissile isotopes. Nuclear resonance fluorescence is one technique that can be used in applications to identify and quantify isotopic content. NRF has also been proposed in several detector systems as a viable way to identify fissile material in cargo containers [2, 3]. It is necessary to measure the characteristic resonances of each isotope that is to be distinguished in identification and assay applications.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2010

Neutron-induced fission cross section of {sup 234}U and {sup 237}Np measured at the CERN Neutron Time-of-Flight (n{sub T}OF) facility

A high-resolution measurement of the neutron-induced fission cross section of {sup 234}U and {sup 237}Np has been performed at the CERN Neutron Time-of-Flight facility. The cross sections have been determined in a wide energy range from 1 eV to 1 GeV using the evaluated {sup 235}U cross section as reference. In these measurements the energy determination for the {sup 234}U resonances could be improved, whereas previous discrepancies for the {sup 237}Np resonances were confirmed. New cross-section data are provided for high neutron energies that go beyond the limits of prior evaluations, obtaining important differences in the case of {sup 237}Np.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2011

MANTRA: An Integral Reactor Physics Experiment to Infer Actinide Capture Cross-sections from Thorium to Californium with Accelerator Mass Spectrometry

The principle of the proposed experiment is to irradiate very pure actinide samples in the Advanced Test Reactor at INL and, after a given time, determine the amount of the different transmutation products. The determination of the nuclide densities before and after neutron irradiation will allow inference of effective neutron capture cross-sections. This approach has been used in the past and the novelty of this experiment is that the atom densities of the different transmutation products will be determined using the Accelerator Mass Spectrometry technique at the ATLAS facility located at ANL. It is currently planned to irradiate the following isotopes: 232Th, 235U, 236U, 238U, 237Np, 238Pu, 239Pu, 240Pu, 241Pu, 242Pu, 241Am, 243Am, 244Cm and 248Cm.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2011

The {sup 237}Np(n,f) cross section at the CERN n-TOF facility

The {sup 237}Np(n,f) cross sections have been measured at the n-TOF facility relative to the {sup 235}U and {sup 238}U fission cross sections. The n-TOF spallation neutron source at CERN is characterized by a high intensity flux, an excellent time resolution and an extensive neutron energy range (from eV to GeV). A fast ionization chamber was used as a fission fragment detector with efficiency better than 97%. Preliminary {sup 237}Np(n,f) cross sections have been deduced in the energy range from 20 keV to 10 MeV and are found in good agreement with data from literature.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2012

Characterization of Delayed-Particle Emission Signatures for Pyroprocessing, Part 1: Fuel Assembly

An initial study was conducted with nondestructive assay analysis using the MCNP6 Monte Carlo radiation-transport code to calculate delayed-neutron and delayed- gamma emission signature monitors to characterize burnup and isotopic composition for pyroprocessing. The study consists of five parts. In Part 1, a generic Westinghouse fuel assembly model was modified to create models with 3% and 5% 235 U enrichments, 20, 30, 40, and 50 GWd/MTU burnups, and 3-, 5-, 10-, 20-, and 30-year cooling times. MCNP6 was used to create detail-rich delayed-particle source terms for fuel assembly. Delayed-gamma activity ratios were then developed for conventional 134 Cs/ 137 Cs, 134 Cs/ 154 Eu, and 154 Eu/ 137 Cs markers as a function of burnup and of actinide mass for elemental uranium, neptunium, plutonium, americium, and curium. In addition, data mining using photon tally data was employed to discover potential new delayed-gamma activity ratios. The data mining effort was streamlined by the creation and use of several utility codes to extract and process data. The study confirms behavior for burnup dependence of 134 Cs/ 137 Cs, 134 Cs/ 154 Eu, and 154 Eu/ 137 Cs markers. A process for the discovery of potential new delayed-gamma markers has been developed, and results are presented. The actinide and nonactinide inventories for the fuel assembly models are used as source materials for the electrorefining models. Parts 2–5 contain analysis of a pyrochemical separations model that includes an MCNP6 mockup of the INL Mark IV electrorefiner (see companion documents). Part 2 focuses on delayed-particle emissions from the electrorefiner with loaded fuel dissolution baskets before electrorefining. Part 3 investigates delayed-particle emissions from the electrorefiner following electrorefining. Part 4 presents delayed-particle emissions from other pyroprocessing units that contain material developed by the electrorefining process in Part 3. Part 5 parallels Part 4, except for a 10 wt% plutonium reduction.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2020

235-F GoldSim Fate and Transport Model: Uncertainty Quantification

Building 235-F was configured with two missions in mind: Actinide Billet Line (ABL) and the fabrication of Pu-238 oxide for space program applications. ABL produced Np-237 billets for use in SRS reactors, whereas the design process, fabrication, and examination of Pu-238 oxide powder occurred within the following areas, respectively: Plutonium Experimental Facility (PEF), Plutonium Fuel Form (PuFF), and Old Metallography Lab (OML). By 1990 production ceased and by 2006 de-inventory occurred; however, assays have shown significant holdup remains within ABL and PuFF. As a result, 235-F is a Category 2 nuclear facility, with plans to undergo deactivation and decommission (D and D) via In-Situ Disposal (ISD). The purpose of this project is to ensure United States Environmental Protection Agency (USEPA) groundwater radiation maximum contaminant level (MCL) and dosage standards are met during the D and D of 235-F by quantifying uncertainty through probabilistic modeling and evaluation of various ISD alternatives. GoldSim is a dynamic modeling software package with a graphical, object-oriented interface capable of capturing the influence of complex system input variability on probabilistic system outcomes. A GoldSim stochastic fate and transport model for 235-F was developed and matched with a PORFLOW deterministic model to simulate probabilistic release and flow of radionuclides from ABL and PuFF into the vadose zone, the Upper Three Runs (UTR) Aquifer, and UTR Creek. The GoldSim model was used to probabilistically evaluate four ISD scenarios against USEPA groundwater radiation MCLs and dosage standards. The deterministic 235-F GoldSim fate and transport model continues to be refined to match the results of the PORFLOW deterministic model to ensure the model accurately represents radionuclide movement through the groundwater system. The stochastic variables that are utilized within the GoldSim model are founded on the most current data; a conservative perspective is taken where needed. Alignment with the PORFLOW deterministic model, coupled with input stochastic variability, allows the probabilistic 235-F GoldSim model to capture the conservative breadth of possible outcomes for radionuclide fate within this particular system. This ensures that the USEPA MCLs and dosage limits hold even in the worst case scenarios.

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

Contaminant Migration Modeling to Support the In Situ Decommissioning of Hardened Nuclear Facilities - 20215

The decommissioning of hardened nuclear facilities provides a unique challenge in balancing current worker dose with exposure to future receptors and media. In situ decommissioning (ISD) is a cost-effective and safe option for the closure of such facilities but requires a detailed understanding of the threat that radioactive and hazardous constituents remaining in the facilities pose to groundwater and surrounding surface water. The Savannah River Site (SRS) extensively employs contaminant migration models within the decommissioning process to develop site-specific removal, grouting, and monitoring strategies in support of safe and effective final end states of nuclear facilities. This work discusses the use of contaminant migration modeling in the ISD process at SRS, including specific examples from the closure of P-Reactor and R-Reactor, Building 235-F, F-Canyon and F/H Laboratory Complex. Migration models developed during the closure of P-Reactor and R-Reactor were divided into four source areas: the reactor vessel, the process area, the disassembly basin, and the purification wing. Each source area was assigned a specific inventory and migration pathway and was then subject to varying hypothetical removal, capping, and grouting schemes. Modeling indicated that groundwater Maximum Concentration Limits (MCLs) might be exceeded if no action was taken for eleven and ten constituents of concern (COCs) at P-Reactor and R-Reactor, respectively. These exceedances could occur in as few as 200 to 500 years. Alternatively, the migration modeling demonstrated that the selected ISD actions reduced contaminant mobility which allowed for significant radioactive decay and resulted in fewer predicted exceedances of groundwater MCLs (five COCs for P-Reactor; eight COCs for R-Reactor). In response to the modeling results for P-Reactor and R-Reactor, effectiveness monitoring programs were developed to target contaminants, specific to each reactor, that may migrate to groundwater. Contaminant migration modeling also revealed that roof collapse was a large factor in the release of COCs to the environment, giving rise to roof improvements, and an inspection and vegetation control program to ensure roof stability over time. Contaminant migration modeling is also aiding in the closure planning for hardened facilities in F Area. Building 235-F housed the Actinide Billet Line, which produced Np-237 billets for irradiation in SRS reactors, and the Plutonium Fuel Form (PuFF) facility that produced Pu-238 heat sources for the space program. As a result of these missions, areas within Building 235-F contain considerable residual amounts of both Pu-238 and Np-237. Contaminant migration modeling of Building 235-F was originally performed in 2012 to identify the feasibility of ISD and the amount of radioactive material removal required to prevent the exceedance of groundwater MCLs. The original model indicated that a 60% reduction in the PuFF facility Pu-238 inventory could keep groundwater concentrations below standards, while Np-237 did not pose a threat to groundwater. However, updates to the model with an emphasis on source impact pathways revealed that, due to the orientation of the source areas relative to groundwater flow, no amount of reasonable removal of Pu-238 would keep groundwater concentrations below MCLs and that Np-237 could be a large contributor to localized MCL exceedances. With this insight, the refined 2019 model is being used to assist in the development of grouting plans specific to each facility source area, where bentonite may be utilized to slow the migration of Pu-238 and its daughter products from the PuFF facility and a reducing grout may decrease Np-237 transport by ensuring the nuclide remains in the less mobile +IV oxidation state. The beginning phases of contaminant migration modeling are underway for F-Canyon and associated facilities using lessons learned from the Reactors and Building 235-F. A contaminant migration pathway, similar to the pathway used for the reactor vessels, is being developed for the hot and warm canyons. The F-Canyon inventories are being spatially refined to identify specific and localized source zones, comparable to Building 235-F, that may impact groundwater. Contaminant migration modeling provides vital input within the ISD process, from facility investigation to post closure effectiveness monitoring, making it a valuable tool in the development of safe and effective end states for hardened nuclear facilities at SRS. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗