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DOE OSTI2016

Data Validation Package November 2015 Groundwater and Surface Water Sampling at the Old and New Rifle, Colorado, Processing Sites February 2016

Water samples were collected from 36 locations at New Rifle and Old Rifle, Colorado, Processing Sites. Duplicate samples were collected from New Rifle locations 0659 and 0855, and Old Rifle location 0304. One equipment blank was collected after decontamination of non-dedicated equipment used to collect one surface water sample. Sampling and analyses were conducted as specified in the Sampling and Analysis Plan for U.S. Department of Energy Office of Legacy Management Sites (LMS/PRO/S04351, continually updated). New Rifle Site Samples were collected at the New Rifle site from 16 monitoring wells and 7 surface locations in compliance with the December 2008 Groundwater Compliance Action Plan [GCAP] for the New Rifle, Colorado, Processing Site (LMS/RFN/S01920), with one exception: New Rifle location 0635 could not be sampled because it was inaccessible; a fence installed by the Colorado Department of Transportation prevents access to this location. DOE is currently negotiating access with the Colorado Department of Transportation. Analytes measured at the New Rifle site included contaminants of concern (COCs) (arsenic, molybdenum, nitrate + nitrite as nitrogen, selenium, uranium, and vanadium) ammonia as nitrogen, major cations, and major anions. Field measurements of total alkalinity, oxidation- reduction potential, pH, specific conductance, turbidity, and temperature were made at each location, and the water level was measured at each sampled well. A proposed alternate concentration limit (ACL) for vanadium of 50 milligrams per liter (mg/L), specific to the compliance (POC) wells (RFN-0217, -0659, -0664, and -0669) is included in the New Rifle GCAP. Vanadium concentrations in the POC wells were below the proposed ACL as shown in the time-concentration graphs in the Data Presentation section (Attachment 2). Time-concentration graphs from all other locations sampled are also included in Attachment 2. Sampling location RFN-0195 was misidentified for the June/August 2014 and November 2014 sampling events. (Well RFN-0609 was inadvertently sampled instead of RFN-0195 in 2014.) The results for RFN-0195 have been corrected, and are included in associated time-concentration graphs for this location. Recent results for RFN-0195 are consistent with established trends with the possible exception of vanadium. The most recent result for vanadium showed an increase over recent values. Vanadium concentrations at RFN-0195 and other locations will continue to be evaluated in the future to determine the potential for deviations from established trends. The surface water locations were sampled to monitor the impact of groundwater discharge. COC concentrations at Colorado River surface water locations RFN-0324 and RFN-0326, downgradient of the site, remained low and were consistent with historical results, as shown in the time-concentration graphs. COC concentrations did not indicate there are any impacts related to groundwater discharge to the river. In many cases, elevated COC concentrations at the New Rifle site pond locations were observed, as shown in the time-versus concentration graphs. As noted in the GCAP, this indicates impacts from groundwater discharge to the ponds. Old Rifle Site Samples were collected at the Old Rifle site from eight monitoring wells and five surface locations in compliance with the December 2001 Groundwater Compliance Action Plan for the Old Rifle, Colorado, UMTRA Project Site (GJ0-2000-177-TAR). Analytes measured at the Old Rifle site included COCs (selenium, uranium, and vanadium), major cations, and major anions. Field measurements of total alkalinity, oxidation-reduction potential, pH, specific conductance, turbidity, temperature, were made at each location, and the water level was measured at each sampled well. The monitoring strategy described in the GCAP is designed to determine progress of the natural flushing process in meeting compliance standards for site COCs. Standards for selenium and vanadium are the proposed ACLs of0.05 mg/L and 1.0 mg/L, respectively. For uranium the cleanup goal is the UMTRA standard of 0.044 mg/L or background, whichever is higher. As shown in the time concentration graphs, the uranium concentration exceeds the cleanup goal at groundwater monitoring locations RF0-0304, -0305, -0310, -0655, and -0656. The surface water locations were sampled to monitor the impact of groundwater discharge at Colorado River surface water locations adjacent to (RF0-0396) and downgradient of the site (RF0-0741). COC concentrations remain low and consistent with historical concentrations as shown in the time-concentration graphs (Attachment 2), which indicate no impacts from groundwater discharge to the river.

2015 Old and New Rifle↗

DOE OSTI2017

Investigation of variable compositions on the removal of technetium from Hanford Waste Treatment Plant low activity waste melter off-gas condensate simulant

The Low Activity Waste (LAW) vitrification facility at the Hanford Waste Treatment and Immobilization Plant (WTP) will generate an aqueous condensate recycle stream (LAW Off-Gas Condensate) from the offgas system. The plan for disposition of this stream during baseline operations is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation and recycled to the LAW vitrification facility again. The primary reason to recycle this stream is so that the semi-volatile 99 Tc isotope eventually becomes incorporated into the glass. This stream also contains non-radioactive salt components that are problematic in the melter, so diversion of this stream to another process would eliminate recycling of these salts and would enable simplified operation of the LAW melter and the Pretreatment Facilities. This diversion from recycling this stream within WTP would have the effect of decreasing the LAW vitrification mission duration and quantity of glass waste. The concept being tested here involves removing the 99 Tc so that the decontaminated aqueous stream, with the problematic salts, can be disposed elsewhere.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2017

Special Analysis: 2016-003 Upgrade of Area G PA=CA Model to Updated Versions of GoldSim Software and to LANL Analysts

The Los Alamos National Laboratory (LANL) generates radioactive waste as a result of various activities. Operational waste is generated from a wide variety of research and development activities including nuclear weapons development, energy production, and medical research. Environmental restoration (ER), and decontamination and decommissioning (D&D) waste is generated as contaminated sites and facilities at LANL undergo cleanup or remediation. The majority of this waste is low-level radioactive waste (LLW) and is disposed of at the Technical Area 54 (TA-54), Area G disposal facility. These analyses estimate rates of radionuclide release from the waste disposed of at the facility, simulate the movement of radionuclides through the environment, and project potential radiation doses to humans for several onsite and offsite exposure scenarios. The assessments are based on existing site and disposal facility data, and assumptions about future rates and methods of waste disposal.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2017

Special Analysis: 2017-001 Disposal of Drums Containing Enriched Uranium in Pit 38 at Technical Area 54, Area G

Los Alamos National Laboratory (LANL) generates radioactive waste as a result of various activities. Operational waste is generated from a wide variety of research and development activities including nuclear weapons development, energy production, and medical research. Environmental restoration (ER), and decontamination and decommissioning (D&D) waste is generated as contaminated sites and facilities at LANL undergo cleanup or remediation. The majority of this waste is low-level radioactive waste (LLW) and is disposed of at the Technical Area 54 (TA-54), Area G disposal facility. This special analysis, SA 2017-001, evaluates the potential impacts of disposing of this waste in Pit 38 at Area G based on the assumptions that form the basis of the Area G PA/CA. Section 2 describes the methods used to conduct the analysis; the results of the evaluation are provided in Section 3; and conclusions and recommendations are provided in Section 4.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2016

FY-2016 Methyl Iodide Higher NOx Adsorption Test Report

Deep-bed methyl iodide adsorption testing has continued in Fiscal Year 2016 under the Department of Energy (DOE) Fuel Cycle Technology (FCT) Program Offgas Sigma Team to further research and advance the technical maturity of solid sorbents for capturing iodine-129 in off-gas streams during used nuclear fuel reprocessing. Adsorption testing with higher levels of NO (approximately 3,300 ppm) and NO2 (up to about 10,000 ppm) indicate that high efficiency iodine capture by silver aerogel remains possible. Maximum iodine decontamination factors (DFs, or the ratio of iodine flowrate in the sorbent bed inlet gas compared to the iodine flowrate in the outlet gas) exceeded 3,000 until bed breakthrough rapidly decreased the DF levels to as low as about 2, when the adsorption capability was near depletion. After breakthrough, nearly all of the uncaptured iodine that remains in the bed outlet gas stream is no longer in the form of the original methyl iodide. The methyl iodide molecules are cleaved in the sorbent bed, even after iodine adsorption is no longer efficient, so that uncaptured iodine is in the form of iodine species soluble in caustic scrubber solutions, and detected and reported here as diatomic I2. The mass transfer zone depths were estimated at 8 inches, somewhat deeper than the 2-5 inch range estimated for both silver aerogels and silver zeolites in prior deep-bed tests, which had lower NOx levels. The maximum iodine adsorption capacity and silver utilization for these higher NOx tests, at about 5-15% of the original sorbent mass, and about 12-35% of the total silver, respectively, were lower than for trends from prior silver aerogel and silver zeolite tests with lower NOx levels. Additional deep-bed testing and analyses are recommended to expand the database for organic iodide adsorption and increase the technical maturity if iodine adsorption processes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Analysis of a Radioactive Release in a Nuclear Waste Disposal Facility

The Waste Isolation Pilot Plant (WIPP) is a 655-meter deep mine near Carlsbad, New Mexico, used to dispose the nation’s defense transuranic waste. Limited airborne radioactivity was released from a container of radioactive waste in WIPP on 14 February, 2014. As designed, a mine ventilation filtration system prevented the large scale release of contamination from the underground. However, isolation dampers leaked, which allowed the release of low levels of contaminants after the event until they were sealed. None of the exposed individuals received any recordable dose. While surface contamination was limited, contamination in the ventilation system and portions of the underground was substantial. High efficiency particulate air (HEPA) filters in the operating ventilation system ensure continued containment during recovery and resumption of disposal operations. However, ventilation flow is restricted since the incident, with all exhaust air directed through the filters. Decontamination and natural fixation by the hygroscopic nature of the salt host rock has reduced the likelihood of further contamination spread. Contamination control and ventilation system operability are crucial for resumption of operations. This article provides an operational assessment and evaluation of these two key areas.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2017

Headspace analysis of foams and fixatives

In the process of decontaminating and decommissioning (D&D) older nuclear facilities, special precautions must be taken with removable or airborne contamination. One possible strategy utilizes foams and fixatives to affix these loose contaminants. Many foams and fixatives are already commercially available, either generically or sold specifically for D&D. However, due to a lack of revelant testing in a radioactive environment, additional verification is needed to confirm that these products not only affix contamination to their surfaces, but also will function in a D&D environment. Several significant safety factors, including flammability and worker safety, can be analyzed through the process of headspace analysis, a technique that analyzes the off gas formed before or during the curing process of the foam/fixative, usually using gas chromatography-mass spectrometry (GC-MS). This process focuses on the volatile components of a chemical, which move freely between the solid/liquid form within the sample and the gaseous form in the area above the sample (the headspace). Between possibly hot conditions in a D&D situation and heat created in a foaming reaction, the volatility of many chemicals can change, and thus different gasses can be released at different times throughout the reaction. This project focused on analysis of volatile chemicals involved in the process of using foams and fixatives to identify any potential hazardous or flammable compounds.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2017

Extraction, scrub, and strip test results for the salt waste processing facility caustic side solvent extraction solvent example

An Extraction, Scrub, and Strip (ESS) test was performed on a sample of Salt Waste Processing Facility (SWPF) Caustic-Side Solvent Extraction (CSSX) solvent and salt simulant to determine cesium distribution ratios (D(Cs)), and cesium concentration in the strip effluent (SE) and decontaminated salt solution (DSS) streams; this data will be used by Parsons to help determine if the solvent is qualified for use at the SWPF. The ESS test showed acceptable performance of the solvent for extraction, scrub, and strip operations. The extraction D(Cs) measured 12.9, exceeding the required value of 8. This value is consistent with results from previous ESS tests using similar solvent formulations. Similarly, scrub and strip cesium distribution ratios fell within acceptable ranges.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2017

Solid secondary waste testing for maintenance of the Hanford Integrated Disposal Facility Performance Assessment - FY 2017

The Waste Treatment and Immobilization Plant (WTP) at Hanford is being constructed to treat 56 million gallons of radioactive waste currently stored in underground tanks at the Hanford site. Operation of the WTP will generate several solid secondary waste (SSW) streams including used process equipment, contaminated tools and instruments, decontamination wastes, high-efficiency particulate air filters (HEPA), carbon adsorption beds, silver mordenite iodine sorbent beds, and spent ion exchange resins (IXr) all of which are to be disposed in the Integrated Disposal Facility (IDF). An applied research and development program was developed using a phased approach to incrementally develop the information necessary to support the IDF PA with each phase of the testing building on results from the previous set of tests and considering new information from the IDF PA calculations. This report contains the results from the exploratory phase, Phase 1 and preliminary results from Phase 2. Phase 3 is expected to begin in the fourth quarter of FY17.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2017

Extraction, scrub, and strip test results for the solvent transfer to salt waste processing facility

The Savannah River National Laboratory (SRNL) prepared approximately 240 gallons of Caustic-Side Solvent Extraction (CSSX) solvent for use at the Salt Waste Processing Facility (SWPF). An Extraction, Scrub, and Strip (ESS) test was performed on a sample of the prepared solvent using a salt solution prepared by Parsons to determine cesium distribution ratios (D(Cs)), and cesium concentration in the strip effluent (SE) and decontaminated salt solution (DSS) streams. This data will be used by Parsons to help qualify the solvent for use at the SWPF. The ESS test showed acceptable performance of the solvent for extraction, scrub, and strip operations. The extraction D(Cs) measured 15.5, exceeding the required value of 8. This value is consistent with results from previous ESS tests using similar solvent formulations. Similarly, scrub and strip cesium distribution ratios fell within acceptable ranges.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2017

Special Analysis: 2016-001 Analysis of the Potential Under-Reporting of Am-241 Inventory for Nitrate Salt Waste at Area G

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2017

Light Water Reactor Sustainability Program, U.S. Efforts in Support of Examinations at Fukushima Daiichi-2017 Evaluations

Although the accident signatures from each unit at the Fukushima Daiichi Nuclear Power Station (NPS) [Daiichi] differ, much is not known about the end-state of core materials within these units. Some of this uncertainty can be attributed to a lack of information related to cooling system operation and cooling water injection. There is also uncertainty in our understanding of phenomena affecting: a) in-vessel core damage progression during severe accidents in boiling water reactors (BWRs), and b) accident progression after vessel failure (ex-vessel progression) for BWRs and Pressurized Water Reactors (PWRs). These uncertainties arise due to limited full scale prototypic data. Similar to what occurred after the accident at Three Mile Island Unit 2, these Daiichi units offer the international community a means to reduce such uncertainties by obtaining prototypic data from multiple full-scale BWR severe accidents. Information obtained from Daiichi is required to inform Decontamination and Decommissioning activities, improving the ability of the Tokyo Electric Power Company Holdings, Incorporated (TEPCO Holdings) to characterize potential hazards and to ensure the safety of workers involved with cleanup activities. This document, which has been updated to include FY2017 information, summarizes results from U.S. efforts to use information obtained by TEPCO Holdings to enhance the safety of existing and future nuclear power plant designs. This effort, which was initiated in 2014 by the Reactor Safety Technologies Pathway of the Department of Energy Office of Nuclear Energy Light Water Reactor (LWR) Sustainability Program, consists of a group of U.S. experts in LWR safety and plant operations that have identified examination needs and are evaluating TEPCO Holdings information from Daiichi that address these needs. Each year, annual reports include examples demonstrating that significant safety insights are being obtained in the areas of component performance, fission product release and transport, debris end-state location, and combustible gas generation and transport. In addition to reducing uncertainties related to severe accident modeling progression, these insights are being used to update guidance for severe accident prevention, mitigation, and emergency planning. Furthermore, reduced uncertainties in modeling the events at Daiichi will improve the realism of reactor safety evaluations and inform future D&D activities by improving the capability for characterizing potential hazards to workers involved with cleanup activities. Highlights in this FY2017 report include new insights with respect to the forces required to produce the observed Daiichi Unit 1 (1F1) shield plug endstate, the observed leakage from 1F1 components, and the amount of combustible gas generation required to produce the observed explosions in Daiichi Units 3 and 4 (1F3 and 1F4). This report contains an appendix with a list of examination needs that was updated after U.S. experts reviewed recently obtained information from examinations at Daiichi. Additional details for higher priority, near-term, examination activities are also provided. This report also includes an appendix with a description of an updated website that has been reformatted to better assist U.S. experts by providing information in an archived retrievable location, as well as an appendix summarizing U.S. Forensics activities to host the TMI-2 Knowledge Transfer and Relevance to Fukushima Meeting that was held in Idaho Falls, ID, on October 10-14, 2016.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2016

Evaluation of Environmental Conditions on the Curing Of Commercial Fixative and Intumescent Coatings

Performance metrics for evaluating commercial fixatives are often not readily available for important parameters that must be considered per the facility safety basis and the facility Basis for Interim Operations (BIO). One such parameter is the behavior of such materials in varied, “non-ideal” conditions where ideal is defined as 75 °F, 40% RH. Coupled with the inherent flammable nature of the fixative materials that can act to propagate flame along surfaces that are otherwise fireproof (concrete, sheet metal), much is left unknown when considering the safety basis implications for introducing these materials into nuclear facilities. Through SRNL’s efforts, three (3) fixatives, one (1) decontamination gel, and six (6) intumescent coatings were examined for their responses to environmental conditions to determine whether these materials were impervious to non-nominal temperatures and humidities that may be found in nuclear facilities. Characteristics that were examined included set-to-touch time, dust free time, and adhesion testing of the fully cured compounds. Of these ten materials, three were two-part epoxy materials while the other seven consisted of only one constituent. The results show that the epoxies tested are unable to cure in sub-freezing temperatures, with the low temperatures inhibiting crosslinking to a very significant degree. These efforts show significant inhibiting of performance for non-nominal environmental conditions, something that must be addressed both in the decision process for a fixative material to apply and per the safety basis to ensure the accurate flammability and material at risk is calculated.

36 MATERIALS SCIENCE↗

DOE OSTI2017

Evaluation of Environmental Conditions on the Curing Of Commercial Fixative and Intumescent Coatings

36 MATERIALS SCIENCE↗

DOE OSTI2017

Extraction, Scrub, and Strip Test Results for the Salt Waste Processing Facility Caustic Side Solvent Extraction Solvent Sample

An Extraction, Scrub, and Strip (ESS) test was performed on a sample of Salt Waste Processing Facility (SWPF) Caustic-Side Solvent Extraction (CSSX) solvent and salt simulant to determine cesium distribution ratios (D( Cs )), and cesium concentration in the strip effluent (SE) and decontaminated salt solution (DSS) streams; this data will be used by Parsons to help determine if the solvent is qualified for use at the SWPF. The ESS test showed acceptable performance of the solvent for extraction, scrub, and strip operations. The extraction D( Cs ) measured 12.5, exceeding the required value of 8. This value is consistent with results from previous ESS tests using similar solvent formulations. Similarly, scrub and strip cesium distribution ratios fell within acceptable ranges. This revision was created to correct an error. The previous revision used an incorrect set of temperature correction coefficients which resulted in slight deviations from the correct D( Cs ) results.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2017

Investigation of Tank 241-AN-101 Floating Solids

Tank 241-AN-101 is the receiver tank for retrieval of several C-Farms waste tanks, including Tanks 241-C-102 and 241-C-111. Tank 241 C 111 received first-cycle decontamination waste from the bismuth phosphate process and Plutonium and Uranium Extraction cladding waste, as well as hydraulic fluid. Three grab samples, 1AN-16-01, 1AN-16-01A, and 1AN-16-01B, were collected at the surface of Tank 241-AN-101 on April 25, 2016, after Tank 241-C-111 retrieval was completed. Floating solids were observed in the three grab samples in the 11A hot cell after the samples were received at the 222-S Laboratory. Routine chemical analyses, solid phase characterization on the floating and settled solids, semivolatile organic analysis mainly on the aqueous phase for identification of degradation products of hydraulic fluids were performed. Investigation of the floating solids is reported.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2017

High efficiency environmental sampling with UV-cured peelable coatings (aka NuGoo project)

This report presents slides on CA Related Project (FY13-17); Environmental sampling by IAEA (not only) during CA; Decontamination gels; Cotton swipes vs. decon gel (FY15); Contamination removal study; The origins of the NuGoo; NuGoo – proof of concept; NuGoo – FY17 project ($250K); LED lamp – which one works and why; Selecting photoinitiator; Monomers and oligomers; Results.

47 OTHER INSTRUMENTATION↗

DOE OSTI2017

Project 57 Air Monitoring Report: January 1 through December 31, 2016

On April 24, 1957, the Atomic Energy Commission (AEC, now the Department of Energy [DOE]) conducted the Project 57 safety experiment in western Emigrant Valley northeast of the Nevada National Security Site (NNSS, formerly the Nevada Test Site) on lands withdrawn by the Department of Defense (DOD) for the Nevada Test and Training Range (NTTR). The test was undertaken to develop (1) a means of estimating plutonium distribution resulting from a non-nuclear detonation; (2) biomedical evaluation techniques for use in plutonium-laden environments; (3) methods of surface decontamination; and (4) instruments and field procedures for prompt estimation of alpha contamination (Shreve, 1958). Although the test did not result in the fission of nuclear materials, it did disseminate plutonium across the land surface. Following the experiment, the AEC fenced the contaminated area and returned control of the surrounding land to the DOD. Various radiological surveys were performed in the area and in 2007, the DOE expanded the demarked Contamination Area (CA) by posting signs 200 to 400 ft (60 to 120 m) outside of the original fence.

54 ENVIRONMENTAL SCIENCES↗