Shannon M. Morley

Liquid scintillation counting of environmental radionuclides: a review of the impact of background reduction

Liquid scintillation counting (LSC) supports a range of environmental science measurements. At Pacific Northwest National Laboratory, we are constructing an LSC system with an expected background reduction of 10–100 relative to values reported in the literature. In this paper, a number of current measurement applications of LSC have been considered with an emphasis on determining which aspects of such measurements would gain the greatest benefit: improved minimum detectable activity (MDA), reduction in sample size, and reduction in total analysis time.

Integrated separation scheme for measuring a suite of fission and activation products from a fresh mixed fission and activation product sample

Mixed fission and activation materials resulting from various nuclear processes and events contain a wide range of isotopes for analysis spanning almost the entire periodic table. This work describes the production of a complex synthetic sample containing fission products, activation products, and irradiated soil, and determines the percent chemical recovery of select isotopes through the integrated chemical separation scheme. Based on the results of this experiment, a complex synthetic sample can be prepared with low atom/fission ratios and isotopes of interest accurately and precisely measured following an integrated chemical separation method.

Development of a low background liquid scintillation counter for a shallow underground laboratory

Pacific Northwest National Laboratory has recently opened a shallow underground laboratory intended for measurement of low-concentration levels of radioactive isotopes in samples collected from the environment. The development of a low-background liquid scintillation counter is currently underway to further augment the measurement capabilities within this underground laboratory. Liquid scintillation counting is especially useful for measuring charged particle (e.g., β and α) emitting isotopes with no (or very weak) gamma-ray yields. The combination of high-efficiency detection of charged particle emission in a liquid scintillation cocktail coupled with the low-background environment of an appropriately designed shield located in a clean underground laboratory provides the opportunity for increased-sensitivity measurements of a range of isotopes. To take advantage of the 35m-water-equivalent overburden of the underground laboratory, a series of simulations have evaluated the scintillation counters shield design requirements to assess the possible background rate achievable. This report presents the design and background evaluation for a shallow underground, low background liquid scintillation counter design for sample measurements.

Optical design considerations for efficient light collection from liquid scintillation counters.

Liquid scintillation counters measure charged particle-emitting radioactive isotopes and are used for environmental studies, nuclear chemistry, and life science. Alpha and beta emissions arising from the material under study interact with the scintillation cocktail to produce light. The prototypical liquid scintillation counter employs low-level photon-counting detectors to measure the arrival of the scintillation. For reliable operation, the counting instrument must convey the scintillation light to the detectors efficiently and predictably. Current best practices employ the use of two or more detectors for coincidence processing to discriminate true scintillation events from background events due to instrumental effects such as photomultiplier tube dark rates, tube flashing, or other light emission not generated in the scintillation cocktail vial. In low-background liquid scintillation counters, additional attention is paid to shielding the scintillation cocktail from naturally occurring radioactive material present in the laboratory and within the instruments construction materials. Low-background design is generally at odds with optimal light collection. This study presents the evolution of a light collection design for liquid scintillation counting (LSC) in a low-background shield. The basic approach to achieve both good light collection and a low-background measurement is described. The baseline signals arising from the scintillation vial are modeled and methods to efficiently collect scintillation light are presented as part of the development of a customized low-background, high-sensitivity LSC system.

Background characterization of an ultra-low background liquid scintillation counter

The Ultra-Low Background Liquid Scintillation Counter developed by Pacific Northwest National Laboratory will expand the application of liquid scintillation counting by enabling lower detection limits and smaller sample volumes. By reducing the overall count rate of the background environment approximately 2 orders of magnitude below that of commercially available systems, backgrounds on the order of tens of counts per day over an energy range of ~3-3600keV can be realized. Initial test results of the ULB LSC show promising results for ultra-low background detection with liquid scintillation counting.

Assay methods for 238U, 232Th, and 210Pb in lead and calibration of 210Bi bremsstrahlung emission from lead

Methods for measuring 238U, 232Th, and 210Pb in refined lead are presented. The 238U and 232Th concentrations are determined using isotope dilution inductively coupled plasma mass spectrometry after anion exchange column separation of dissolved lead samples. The 210Pb concentration is inferred through α-spectroscopy of a daughter isotope, 210Po, after precipitation separation of dissolved lead samples. Subsequent to the 210Po α-spectroscopy measurement, a method for evaluating 210Pb concentrations was developed via measurement of bremsstrahlung radiation from β-decay of a daughter isotope, 210Bi, using a 14-crystal array of high purity germanium detectors. Ten sources of refined lead were assayed and results are presented.

PNNL Measurement Results for the 2016 Criticality Accident Dosimetry Exercise at the Nevada National Security Stite (IER-148)

iii Summary v Acknowledgments vii Acronyms and Abbreviations ix 1.0 Introduction 1 2.0 Dosimeter Descriptions 3 2.1 Hanford PNAD 3 2.2 PNNL PNAD 6 2.3 PNNL FNAD 7 2.4 Electronic Personal Dosimeters 11 3.0 Experimental Design 12 4.0 Activity Measurements 14 4.1 Sulfur Pellets (P) 15 4.2 Copper Foils (Cu) 19 4.3 Indium Foils (In, In) 21 4.4 Gold Foils (Au) 22 4.5 Specific Activity per Unit Fluence 22 5.0 Methodology for Calculating Neutron Dose from Foil Activity 23 5.1 Neutron Fluence Calculation 23 5.2 Neutron Dose Calculation 26 6.0 Neutron Dose Results Calculated from Foil Activity 27 7.0 InLight® OSL/OSLN Measurements 29 7.1 microStar® Reader Calibration 30 7.2 InLight® Dose Calculation 31 8.0 nanoDot OSL/OSLN Measurements 33 8.1 nanoDot Gamma Dose Measurement in the Hanford PNAD 33 8.2 nanoDot Gamma and Neutron Dose Measurement in the FNAD 34 9.0 Dosimeter Performance Evaluation 35 10.0 Electronic Personal Dosimeter Measurements 36 11.0 Quick Sort Measurements 37 11.1 Direct Survey Measurements on BOMAB Phantoms 37 11.2 Direct Survey Measurements on PNADs 37 12.0 Discussion 38 13.0 Conclusions 40 14.0 References 41

Activation product analysis in a mixed sample containing both fission and neutron activation products

This work describes a radiochemical separation procedure for the determination of gold (Au), platinum (Pt), tantalum (Ta), and tungsten (W) activation in the presence of fission products. Chemical separations result in a reduction in the minimum detectable activity by a factor of 287, 207, 141, and 471 for 182Ta, 187W, 197Pt, and 198Au respectively, with greater than 90% recovery for all elements. These results represent the highest recoveries and lowest minimum detectable activities for 182Ta, 187W, 197Pt, and 198Au from mixed fission-activation product samples to date, enabling considerable refinement in the measurement uncertainties for neutron fluences in highly complex sample matrices.Graphical Abstract

Analysis of fuel using the Direct LSC method determination of bio-originated fuel in the presence of quenching

A modified version of the Direct LSC method to correct for quenching effect was investigated for the determination of bio-originated fuel content in fuel samples produced from multiple biological starting materials. The modified method was found to be accurate in determining the percent bio-originated fuel to within 5% of the actual value for samples with quenching effects ≤43%. Analysis of highly quenched samples was possible when diluted with the exception of one sample with a 100% quenching effect.

Joint IAEA / NNSA International Workshop Nuclear Forensics Methodologies for Practitioners 2013 Scenario Based Exercise – Version 4.0 Instructor’s Manual

[Participants will serve as border guards for Reimerland. They will be given brief instruction on the operation of hand‐held RadioIsotope DetectorS (RIDS) and be provided an intelligence briefing that tells them to be on the lookout for suspicious activity at their post. Their instruction will include directing suspicious vehicles to a location for secondary screening. If, after secondary screening, suspicions of a criminal act involving nuclear and or radioactive materials remain, participants have been instructed to request assistance from the NLEA, who will then setup and manage a radiological crime scene. Participants will watch a demonstration of two vehicles containing radioactive materials driving through and setting off a portal monitor. The first vehicle, a semi‐tractor trailer, sets off only a gamma alarm. After the driver provides a shipping manifest of fertilizer, participants, posing as border guards, are expected to waive this vehicle through inspection. The second vehicle, an SUV, set off both gamma and 2 neutron alarms. The alarming of the neutron monitor should prompt participants to set up a secondary inspection of the vehicle immediately. The driver of the vehicle indicates he is in legal possession of an industrial instrument containing an old 133Ba source that has decayed to a level no longer requiring official paperwork according to the IAEA and internationally accepted transportation regulations. Authorities have verified that the industrial source does fit the description of one that is sold commercially. However, upon setting up a secondary screening, participants will use hand‐held detectors to locate several other radioactive sources emanating from a black duffle bag in the rear of the vehicle (Figure 1). Hand held detectors detect the presence of 133Ba, and Pu. Upon questioning, the driver only commits to having the 133Ba industrial source and cannot account for the detection of neutrons within his vehicle. Since neutron alarms also sounded, participants should indicate that a neutron alarm would be inconsistent with a 133Ba source alone and should therefore conclude further investigation is warranted. This will prompt participants to call in a response team from the NLEA to set up a radiological crime scene around the vehicle in question. The response team is able to shoot a 3‐D X‐ray radiograph of the duffle bag without moving it to ensure it is rendered safe and moveable without disturbing the contents in the field (Figure 2). At this point, the duffle bag is entered into inventory as evidence and a chain of custody form is initiated. Swipes are taken from the outer bag to confirm there is no dispersible contamination. The bag and its contents are considered valuable for the investigation by the lead investigator. He determines the duffle bag is safe to transport to RRL for evidence inventory and analysis. The duffle bag and its contents are packaged and sent off to the RRL.]