CHARACTERIZATION OF FRACKING SOIL, SEDIMENT, AND ...

The Pennsylvania State University

The Graduate School

Department of Mechanical and Nuclear Engineering

CHARACTERIZATION OF FRACKING SOIL, SEDIMENT, AND WASTEWATER

SAMPLES USING

COMPARATIVE NEUTRON ACTIVATION ANALYSIS METHOD

A Thesis in

Nuclear Engineering

by

Maksat Kuatbek

© 2018 Maksat Kuatbek

Submitted in Partial Fulfillment

of the Requirements

for the Degree of

Master of Science

December 2018

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The thesis of Maksat Kuatbek was reviewed and approved* by the following:

Kenan Ünlü

Professor of Nuclear Engineering

Director of Radiation Science and Engineering Center

Thesis Co-Advisor

Amanda M Johnsen

Assistant Research Professor, Radiation Science and Engineering Center

Thesis Co-Advisor

Arthur Thompson Motta

Professor of Nuclear Engineering and Material Science and Engineering

Chair of the Nuclear Engineering Program

*Signatures are on file in the Graduate School

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ABSTRACT

The accurate multi-elemental analysis of soil, sediment, and wastewater samples is

extremely important for the regulatory monitoring of oil and gas (O&G) development. This kind

of analysis is typically conducted using several conventional methods, such as inductively coupled

plasma optical emission spectrometry (ICP-OES) or mass spectrometry (ICP-MS). The main

objective of this study was to apply the neutron activation analysis (NAA) method for qualitative

and quantitative analysis of hydraulic fracturing samples and to evaluate its accuracy and

applicability.

In this work, seventeen solid (soil and sediment) and five liquid (wastewater) samples

collected from the wellbores within Pennsylvania were characterized. The analysis was conducted

at the Radiation Science and Engineering Center (RSEC) using the comparative neutron analysis

method. The Montana II Soil and Buffalo River Sediment certified standard reference materials

obtained from the National Institute of Standards and Technology were used as the comparators.

As the result of this research, the concentration of short-, intermediate-, and long-lived isotopes of

Cl, Mn, Eu, K, Na, As, La, Ca, Ba, Rb, Pa (thorium activation product), Cr, Fe, Hg, Sr, Sc, Se, Zn,

and Cs elements in fracking samples were determined with an accuracy of ppm (mg/g or mg/l).

The experimentally measured values then were analyzed for standard deviation and verified

through a quality control check, with the exception of cesium and chromium; thus, their values

were declared as non-certified.

The trace element concentration values of three oil and gas wastewater samples, which

were obtained by the Comparative NAA (CNAA), were compared with the most probable value

(MPV) results determined via an inter-laboratory study. The MPVs were evaluated using the

nonparametric statistical method on the results collected from 15 laboratories from the United

States, Canada, and Germany that used different equipment and techniques for wastewater

characterization. The comparison results were demonstrated in the percentage difference

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magnitudes that vary from 0.1% to 56.6%. There are several possible reasons that might cause such

a relatively high error, as the hydride concentration remained after dehydration, the mass error of

liquid samples (due to evaporation), the use of multiple vials during dehydration (sample

movement), and the fragmentation of target elements (due to unfulfillment of pulverization and

homogenization of dried crystals before sampling).

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TABLE OF CONTENTS

List of Figures .......................................................................................................................... vii

List of Tables ........................................................................................................................... ix

Acknowledgements .................................................................................................................. xii

Chapter 1 Introduction ............................................................................................................. 1

1.1 Motivation and Objectives ........................................................................................ 1 1.2 Thesis Structure .......................................................................................................... 2

Chapter 2 Hydraulic Fracturing ............................................................................................... 4

2.1 Description of The Fracking Process ......................................................................... 5

2.2 Environmental impacts and Potential Risks ............................................................... 8

Chapter 3 Neutron Activation Analysis ................................................................................... 10

3.1 Background and Specifications of NAA ................................................................... 10 3.2 Neutron Interactions with Matter ............................................................................... 11 3.3 NAA Methods ........................................................................................................... 13 3.3.1 Single Comparator NAA .................................................................................... 13

3.3.2 Instrumental NAA .............................................................................................. 14 3.3.3 Comparative NAA ............................................................................................. 16 3.4 Applicability of NAA ................................................................................................. 17

Chapter 4 Radiation Science and Engineering Center NAA Facility ...................................... 19

4.1 The Penn State Breazeale Nuclear Reactor (PSBR) .................................................. 19 4.2 RSEC Radionuclear Applications Laboratory ........................................................... 21 4.2.1 The Automatic Sample Handling System (ASHS) ............................................ 21

4.2.2 The Counting System ......................................................................................... 22 4.3 PSBR Irradiation Fixtures .......................................................................................... 25

4.4 Neutron Flux Characterization of the Dry Tube 1 for Core 58 Loading .................... 27

4.4.1 Preparation of the Samples and Documentations .............................................. 29

4.4.2 Irradiation and Counting of the Samples ........................................................... 30

4.4.3 Analysis of the Collected Data .......................................................................... 32

Chapter 5 Fracking Soil, Sediment and Wastewater Samples ................................................. 38

Chapter 6 The Experiment ....................................................................................................... 41

6.1 Activity Prediction ..................................................................................................... 41 6.2 The Sample Preparation ............................................................................................. 42

6.3 The Sample Irradiation and Counting ........................................................................ 49

Chapter 7 Experimental Results ............................................................................................... 54

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7.1 Data Analysis ............................................................................................................. 54 7.2 Quality Control .......................................................................................................... 60 7.3 Interlabratory Comparison of Results ........................................................................ 62

Chapter 8 Conclusion and Future Works ................................................................................. 68

References ................................................................................................................................ 73

Appendix A Physical Specifications ................................................................................ 76

Appendix B NIST Certificates ......................................................................................... 77 Appendix C Activity Prediction Results .......................................................................... 84

Appendix D Analysis Results .......................................................................................... 89

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LIST OF FIGURES

Figure 2.1. Comparison of Well Sites... ................................................................................... 4

Figure 2.1.1. Map of shale gas basins in the USA. .................................................................. 5

Figure 2.1.2. Volumetric percentage of additives in fracking fluids ....................................... 7

Figure 3.2.1. Schematic representation of radioactive capture reaction .................................. 12

Figure 4.1.1. A map of the PSBR Core 58 Loading ................................................................ 20

Figure 4.2.1.1. The rotary sample holder of the ASHS (capacity is over 90 samples) ............ 22

Figure 4.2.2.1. Component diagram of instrumentation layout to perform automated

radiation counting .................................................................................................................... 23

Figure 4.3.1 Shape and design of the dry tubes (DT) .............................................................. 26

Figure 4.3.2 The terminus located in the Radionuclear Applications Laboratory (RAL). ...... 27

Figure 4.4.1.1. The cadmium covered wire positions within Dry Tube 1 with respect to fuel

rod ............................................................................................................................................ 30

Figure 4.4.3.1. The AR (Activity Ratio) within the DT1 (Dry Tube 1) ................................... 34

Figure 4.4.3.2. Measured thermal neutron flux within DT1 (Dry Tube 1). ............................. 35

Figure 4.4.3.3. Measured resonance neutron flux within DT1 (Dry Tube 1). ......................... 36

Figure 4.4.3.4. The thermal and resonance neutron flux peak positions within the DT1 in

regard to a PSBR fuel rod ........................................................................................................ 37

Figure 5.1. A picture of all tested fracking soil, sediment, and wastewater samples in their

original plastic containers.. ...................................................................................................... 38

Figure 6.1.1. The graphical user interface (GUI) of the Activity Prediction Tool.. ................. 42

Figure 6.2.1. The Se sample with a concentration of 100 ppm (on the left) and a standard

PTTS capsule with the Se sample loading (on the right).. ....................................................... 45

Figure 6.2.2. A picture of fracking and SRM samples placed into Bucket #1. ........................ 48

Figure 6.2.3. A picture of fracking and SRM samples placed into Bucket #2. ........................ 48

Figure 6.2.4. The aluminum bucket dimensions and sample loading patterns.. ...................... 49

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Figure 6.3.1. The workplace that was set up near the shadow shield corner. It was used to

prepare irradiated samples for gamma ray counting ................................................................ 51

Figure 7.1.1. A gamma spectrum obtained from counting Solid 01 Sample after the

'medium' decay period ............................................................................................................. 57

Figure 7.1.2. A gamma spectrum obtained from counting Liquid 01 Sample after the

'medium' decay period ............................................................................................................. 58

Figure 7.3.1. A comparison of manganese concentrations in oil and gas wastewater samples

in a graphical manner ............................................................................................................... 66

Figure A-1. The HPGe detector dimensions provided by the manufacturer ............................ 76

Figure D-1. A graphical comparison of sodium concentrations in oil and gas wastewater

samples... .................................................................................................................................. 100

Figure D-2. A graphical comparison of potassium concentrations in oil and gas

wastewater samples... ............................................................................................................... 100

Figure D-3. A graphical comparison of calcium concentrations in oil and gas wastewater

samples... .................................................................................................................................. 101

Figure D-4. A graphical comparison of strontium concentrations in oil and gas

wastewater samples... ............................................................................................................... 101

Figure D-5. A graphical comparison of barium concentrations in oil and gas wastewater

samples... .................................................................................................................................. 102

Figure D-6. A graphical comparison of iron concentrations in oil and gas wastewater

samples... .................................................................................................................................. 102

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LIST OF TABLES

Table 2.1.1. Chemical content of a fracturing fluid and specific purposes for hydraulic

fracturing operations ................................................................................................................ 7

Table 4.2.2.1. The outline of the spectrum analysis sequence steps and their purposes. ......... 24

Table 5.1. List of the analyzed samples ................................................................................... 39

Table 6.2.1. Typical trace impurities in Heraeus Suprasil 310 quartz glass ............................ 43

Table 6.2.2. Comparison of selenium saturation activities within original and evaporated

samples. .................................................................................................................................... 46

Table 6.2.3. The identification numbers and weights of the test samples................................ 47

Table 7.1.1. The list of elements of interest and their radionuclides with gamma-decay

energies used in this study ....................................................................................................... 55

Table 7.1.2. Trace element concentrations of Solid 01 Sample ............................................... 56

Table 7.1.3. Trace element concentrations of the fracking samples (Part 1). The values are

given in weight percent (wt%) ................................................................................................. 59

Table 7.1.4. Trace element concentrations of the fracking samples (Part 2). The values are

given in weight percent (wt%) ................................................................................................. 60

Table 7.2.1. A summary of quality control analysis for Bucket#1 .......................................... 62

Table 7.2.2. A summary of quality control analysis for Bucket#2. ......................................... 62

Table 7.3.1. A summary of inter-laboratory study. All values are represented in mg/l ........... 65

Table 7.3.2. The concentration and standard deviation values of some trace elements

measured using the NAA method. ........................................................................................... 65

Table 7.3.3. The percent difference magnitudes between MPV and NAA measured values.

................................................................................................................................................. 67

Table C-1. Calculated activities and dose rates for the end of short irradiation of Bucket #1

content after a decay period of 48 hours .................................................................................. 84


content after a decay period of 48 hours .................................................................................. 86


and Bucket #2 content after a decay period of 192 hours ........................................................ 88

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and Bucket #2 content after a decay period of 552 hours ........................................................ 89

Table D-1. Experimentally determined trace element concentrations of HR SaH sample

using NAA method .................................................................................................................. 89

Table D-2. Experimentally determined trace element concentrations of BO1 sample using

NAA method ............................................................................................................................ 90


NAA method ............................................................................................................................ 90

Table D-4. Experimentally determined trace element concentrations of Sample 02 Solid

sample using NAA method ...................................................................................................... 91


sample using NAA method. ..................................................................................................... 91



Table D-7. Experimentally determined trace element concentrations of AMD cycle 2


Table D-8. Experimentally determined trace element concentrations of AMD test 5 sample

using NAA method .................................................................................................................. 93

Table D-9. Experimentally determined trace element concentrations of AMD test 6

sample using NAA method. ..................................................................................................... 93

Table D-10. Experimentally determined trace element concentrations of HR Evop. Test 01


Table D-11. Experimentally determined trace element concentrations of HR Evop. Test

02 sample using NAA method ................................................................................................. 94

Table D-12. Experimentally determined trace element concentrations of Raw flowhart

solid sample using NAA method ............................................................................................. 95

Table D-13. Experimentally determined trace element concentrations of FS3 Effluent


Table D-14. Experimentally determined trace element concentrations of Marsellus

Flowback sample using NAA method. .................................................................................... 96

Table D-15. Experimentally determined trace element concentrations of Franklin

discharge sample using NAA method ...................................................................................... 96

Table D-16. Experimentally determined trace element concentrations of Sample 01 liquid


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03 sample using NAA method. ................................................................................................ 99

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ACKNOWLEDGEMENTS

The following people played an important role in helping me to achieve everything I have

achieved by this point. That is why in this acknowledgment I want to express my sincere

thankfulness for all their support.

First of all, I would like to thank my co-advisers Professor Kenan Ünlü and Dr. Amanda

Johnsen for their guidance throughout my studies, for their help in and out of academic frames and

most importantly for their motivational support. I acknowledge and appreciate their supervision

and assistance; thus, I can gladly say that for me it was a privilege to work with them. I also thank

my peers and friends from office: Alibek Kenges, Gokhan Corak, Bryan Eyers, Adam Rau, Andrew

Bascom, and Colleen Mulhollan for going through essential discussions together that enhanced my

understanding in all conceptual questions of my thesis. I am grateful to the personnel of the

Radiation Science and Engineering Center, especially Brian Bennett, for allowing me to use his

shop for many hours while producing quartz ampoules.

Moreover, I am thankful to Dr. Nathaniel R. Warner and his student Travis Tasker for

providing us with test samples and the opportunity to become part of the inter-laboratory study.

Finally, and most importantly I thank my family: my parents - Zharylkasyn and Roza for

upbringing me a way that I became a person who always strives for improvement, my sisters for

believing in me and supporting me from far distance with their warm, and all my friends for being

my comrades in all ideas and beliefs.

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Chapter 1

Introduction

1.1 Motivation and Objectives

The accurate multi-elemental analysis of wastewater, soil, and sediment samples is

extremely important for the regulatory monitoring of oil and gas (O&G) development. The output

of this analysis ensures that the concentration of constituent elements within those samples is not

exceeding the regulatory limits. Moreover, in case of potential contamination events, the results

can be used as a fingerprinting application for identifying O&G wastewaters and their headsprings

[1]; thus, the analysis must be very precise and carry both qualitative and quantitative characters.

The chemical characterization of O&G samples is challenging due to the complex sample

(solid and fluid) matrix. It is typically conducted using numerous techniques, such as inductively

coupled plasma with optical emission spectrometry (ICP-OES), inductively coupled plasma with

mass spectrometry (ICP-MS), direct plasma spectrometry (DCP), triple quadrupole inductively

coupled plasma with mass spectrometry (ICP-MS/MS), X-ray fluorescence (XRF), and ion

chromatography (IC). [1]. However, all these methods have limitations due to the sample matrix

specifications or certain deficiencies in the techniques. For instance, ICP-OES or ICP-MS can be

hampered for detecting metals in high salinity O&G wastewaters because of the signal suppression

caused by easily ionized elements such as Na and K. This matrix effect can be solved by simple

sample dilution, which can cause another problem associated with a lack of sensitivity and

exceeding the detection limits for trace metals of interest [1].

The motivation of this study is to apply the neutron activation analysis (NAA) method for

multi-elemental characterization of solid and liquid samples from fracking process. The

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applicability of the NAA for this study will be evaluated by comparing its results with the results

of other methods. Moreover, this comparison allows the determination of the advantages and

disadvantages of the NAA, among other techniques, and to identify the elements for which the

NAA can provide reliable quantitative results. If NAA performs better on some or most of the

elements, it has the full potential to become a secondary or even primary application for O&G

regulation studies.

1.2 Thesis Structure

This thesis consists of eight chapters. This chapter will cover a brief description of the main

objectives and content of each part.

Chapter two discusses a quick overview of the history and future of hydraulic fracturing.

Moreover, the chapter provides a basic knowledge of the fracking process and its methodology.

The types and specifics of fracking fluids, their environmental impacts, and potential risks also

covered in this chapter.

Chapter three provides the background information about the neutron activation analysis

(NAA) technique and the theory behind it. The chapter also introduces the types of neutron

interactions with matter and the most common three NAA methods, such as the absolute NAA

method, comparative method, and single comparator NAA (known as k-factor method). At the end

of this chapter, the applicability of NAA is discussed through existing NAA applications in a wide

range of disciplines.

The fourth chapter describes the Radiation Science and Engineering Center (RSEC)

facilities employed in this research. In addition, this chapter provides a more detailed description

of the automatic sample handling system (ASHS), radiation counting system, the pneumatic tube

transport system (PTTS), and dry irradiation tubes. The detailed data on the neutron flux

characterization of the Dry Tube 1 for Core 58 loading are also given in this chapter.

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A list of analyzed fracking solid (soil and sediment) and liquid (wastewater) samples are

presented in Chapter five. The chapter also provides brief descriptions and specifications of test

samples.

The experimental procedures for this study were divided into three stages and are described

in detail in Chapter 6. The first stage is activity prediction, which was performed using the Activity

Predictor program developed by Dr. Dağistan Şahin. At this step a rough design of the experiment

was determined regarding the choice of the suitable irradiation time and fixture (due to neutron flux

requirements), test sample mass and geometry, time of decay and radiation counting. The second

stage is sample preparation, which mainly describes the test sample weighing and encapsulating

activities. The last stage is the sample irradiation and counting that contains detailed information

on the conditions of irradiation and measurement.

Chapter seven presents the experimental results obtained through comprehensive data

acquisition and analysis. Furthermore, this chapter numerically demonstrates the quality control of

results and their comparison with the results of other laboratories and methodologies.

Chapter eight provides conclusions and a brief summary of this research. Further studies

and work suggestions are also commented on this chapter.

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Chapter 2

Hydraulic Fracturing

Historically, gas-well drilling was performed using a single vertical well, which provided

access to conventional sources of gas that flowed through pore spaces along the wellbore (Figure

2.1). However, there are also unconventional gas reservoirs with low permeability formations that

require a different extraction technique. [2]. For this reason, a new method called “Hydraulic

fracturing”, commonly known as “fracking” was developed. This method allows extraction of

natural gas from deep shale strata by using a high-pressure drilling technique. The drilling process

combines the traditional vertical and additional horizontal drillings that allows injection of highly

pressurized fracking fluid into shale formation to keep fractures propped open, so gas can be

released and freely flow at a higher rate to the wellbore (Figure 2.1) [3] [4].

Figure 2.1. Comparison of Well Sites [5].

Even though the hydrofracking was first used in the 1940s, in practice, it was widely

applied only after the 1990s, when natural gas prices increased making fracking more financially

attractive [2]. Moreover, the latest advances and improvements in horizontal drilling, such as

multiple wells drilled from one surface location, have made this method even more productive and

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economically competitive. Thus, in the last two decades, the number of natural gas wells in the

United States has increased by 200,000, which will allow gas production rate up to 1065 billion

cubic meters per year by 2040 [6]. This scale of production makes hydraulic fracturing a promising

new energy extraction technology of the 21st century [4].

2.1 Description of The Fracking Process

After detailed geological research of deep underground rock formations, the fracking

process starts with drilling and installing wells. A typical installation contains one to several wells

that are drilled from a single wellpad [7]. The fracturing depth depends on target shale stratum, so

major U.S. wells descend vertically from 150 m to more than 4000 m [8]. Figure 2.1.1 shows a

map of shale gas basins in the United States, which are separated and differently colored depending

on their depth and age. The map was prepared by Cidney Christie (Duke University), based on data

from U.S. Energy Information Administration (EIA).

Figure 2.1.1. Map of shale gas basins in the USA [9].

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The horizontal leg of a gas well might continue as much as 1.5 km with discrete length

fractures of 91-152 m (Figure 2.1). In other words, a single horizontal well can allow up to 15

separate hydrofrack “events” simultaneously [2]. The rest of the hydraulic fracturing process can

be explained by three major steps: 1) fracture the rock formation by injection of fracking fluid

(water, sand, and chemical additives) to horizontal drilled well using high pressure; 2) extract and

collect the natural gas released from the shale through the well; 3) treat or dispose of the water that

was used for well fracking. Thus, one of the biggest challenges is the significant volumes of

ascended water that occurs after pressure release. The performance of hydraulic fracturing per well

requires about 2-5 million gallons of water [4]. Around 10 to 80 percent of the injected water may

return to the surface as wastewater. In the entire fracking process, there are two terms regarding

wastewater: flowback (fluid that quickly returns to the surface) and produced water (fluid that takes

longer to return to the surface). Since the injected fluid allows for the liberation of gas and materials

trapped in the shale, flowback and productive water are enriched with brines, hydrocarbons,

naturally occurring radioactive materials, and trace elements. In fact, the longer the fluid remains

in the shale the greater the concentration of native geological formation materials in it [10].

The content on the fracturing fluid varies, depending on the specific needs of the extracting

company and the geological characteristics of the fracking location. Moreover, the inventory and

exact concentration of chemical additives in fracking waters remains confidential; however, it is

possible to roughly evaluate the percentage of volumes of fracking fluid by using general

knowledge about fracking basics and widely reported documents. Overall, the concentration of

different chemical additives in vast fracking fluids is relatively small and vary between 0.5-2%, so

the remaining 98-99.5% of fluid contain water and proppants (silica sand) [5]. The typical

volumetric percentage of additives that were used for a regular fracking treatment is demonstrated

in figure 2.1.2.

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Figure 2.1.2. Volumetric Percentage of Additives in Fracking Fluids [4].

Table 2.1.1 shows the list of major additives, their chemical composition, volumetric

percentage, and a brief explanation regarding a specific purpose of usage.

Table 2.1.1 Chemical content of a fracturing fluid and specific purposes for hydraulic

fracturing operations [4].

However, wastewater (especially produced water) may contain an even wider range of

isotopes than was originally added to the injected water. This phenomenon occurs due to the

mixture of injected water with naturally occurring water in the geologic shale. Thus, the natural

components of the formation will be present in the wastewater when it is recovered from a well

after a long-term period. These substances can originate from the water, rock, oil or gas present in

the formation [4]. Below are described a couple of the most significant classes of constituents:

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• Naturally occurring radioactive materials, such as uranium, thorium, radium, radon,

strontium and potassium [4]. Some research results demonstrate that flowback and produced

water samples contain relatively high concentration of radioactive material. For example, in

a produced water sample from the Marcellus Shale showed radium and uranium

concentrations at the pCi/L level [11]. Similarly, another study identified radium in produced

waters from the Northern Appalachian Basin [4].

• Inorganic substances and metals, such as aluminum, arsenic, barium, bromine, cadmium,

chloride, chromium, iron, manganese, mercury, nickel, sodium, vanadium and zinc. By

nature, the salinity of formation water is very high, so Cl, Na, and Br are the most common

detected elements. Moreover, there are also a list of regularly detected elements from

produced water samples. For example, Ba, Br Ca, Cl, Na, and Sr were common for the

Marcellus Shale basin waters [4].

2.2 Environmental Impacts and Potential Risks

Even though the volumetric percentage of additives in the injected water is very small, the

total amount of chemicals is still significant due to the volume of used water (up to 5 million gallons

per well). Regardless of the level of dilution, this amount of chemicals might carry a potential risk

to the local environment and public health. This fact raised public concerns about fracking, so

scientists started to collect data, analyze, and evaluate the potential risks based on local

environmental impacts. The major potential risks are related to water contamination, air pollution

(large and high-density gas emission), seismicity (small earthquakes), and local landscape changes

[3].

The biggest concern was regarding the potential contamination of water resources, which

includes: 1) stray gas contamination of shallow groundwater that located above shale gas basins;

2) the contamination of pathways and hydraulic connections between shallow drinking water

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aquifers and the deep shale gas formations; 3) inadequate treatment or disposal of wastewater

(flowback and produced water) that causes surface water contamination and long-term ecological

effect [9]. In the case with wastewater disposal everything is relatively straight forward, since this

problem can be regulated by developing new policies. For example, on May 19, 2011 the

Pennsylvania Department of Environmental Protection (PADEP) prohibited to drilling companies

to dispose their wastewater through wastewater treatment plants (WWTPs) [10]. Nevertheless, it is

more challenging to predict and prevent the shallow water aquifer pollution, since contaminants

can potentially be transported through both bulk media (advective way) and fractures (preferential

flow). Moreover, there is a significant proof that the natural vertical flow also drives contaminants

(mostly brine) close to the surface from deep evaporate sources [12]. Thus, it is even more

challenging to sensitively distinguish the contaminations caused by hydraulic fracturing activities

from the pollution due to the natural flow. Recent studies conducted in Marcellus Formation have

shown that strontium isotope ratios (87Sr/86Sr) can be used to investigate the origin of total dissolved

solids (TDS) in ground and surface water [13].

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Chapter 3

Neutron Activation Analysis

Neutron Activation Analysis (NAA) is a very sensitive, non-destructive analytical method

for determining the major, minor, and trace elements of a sample material. Moreover, this technique

allows both qualitative and quantitative multielement analysis. Within proper conditions, NAA is

capable to quantitively identify up to 60 elements in small samples, usually with masses of

milligram. The lower detection limit of NAA varies due to the element or isotope of interest and

typically ranges on the order of parts per million (ppm) to parts per billion (ppb).

3.1 Background and Specifications of NAA

Neutron Activation Analysis was discovered in 1936 by George de Hevesy and Hilde Levi,

when they were performing a quantitative analysis on rare-earth salts by exposing them with

neutrons naturally emitted from Ra(Be) source [14]. In the 1950s and 1960s, the potential of NAA

as an experimental method drastically increased due to more detailed research on the notions such

as, decay, characteristics of radiation absorption, and radiochemical separation. Also, the

introduction of scintillator and semiconductor detectors provided selectivity in gamma-ray

spectrometry, so the individual radionuclides can be identified mostly without initial chemical

separations [15]. The principal involved in NAA consists of bombarding the specimen with

neutrons in a suitable irradiation facility (typically a nuclear research reactor) to produce specific

radionuclides. Following irradiation, the artificially created radionuclides undergo decay to reach

their ground state configurations by emitting beta particles and characteristic gamma rays. Because

each radioactive isotope always emits characteristic gamma rays at unique energies and intensities,

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the quantitative measurement of those gamma rays by gamma spectroscopy provides information

on the radioisotopes present, and hence the parent chemical element(s).

3.2 Neutron Interactions with Matter

Neutrons are electrically neutral, so when they interact with matter, they cannot be affected

by the Coulomb force of either the atomic electron cloud of an atom or the positively charged

nucleus. Therefore, neutrons do not interact with the atom, but directly with the nucleus. The

probability that a neutron interacts with a nucleus is quantitatively described by the term known as

cross-section. The interaction of a neutron with a nucleus may follow one or more of these

reactions: elastic scattering, inelastic scattering, radiative capture, charged-particle reactions,

neutron-producing reactions, and fission [16]. Each reaction has its own cross-sectional value, and

their sum is equal to the total cross-section. Some of these reactions are defined below.

Elastic scattering is when a neutron collides with a nucleus of an atom without changing

its intrinsic composition (number of neutrons and protons) and energy level. All of the kinetic

energy of the incoming neutron is divided between two particles, so after the collision, they recoil

from each other with different speeds and directions. Thus, in the elastic scattering, the nucleus

remains in the initial ground state, despite the energy transfer. In the notation of nuclear reactions,

this reaction is commonly abbreviated as (n,n), demonstrating that the neutron interaction has not

caused any fundamental changes to the nucleus.

Inelastic scattering is a similar process to elastic scattering, except that some portion of the

kinetic energy retained by the nucleus converts to internal energy (an endothermic interaction), so

the nucleus moves from the ground state to the excited state. The excited nucleus eventually decays,

emitting gamma ray(s) (inelastic 𝛾-ray(s)) and returning to its initial ground state. Inelastic

scattering is typically denoted by (n,n’) symbol.

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Radiative capture or neutron absorption is a reaction when the nucleus captures the

colliding neutron and changes its mass and energy. The probability of this event is described by the

neutron capture cross-section, which varies with respect to the size and stability of the target

nucleus, and the energy of incident neutron. After the capture, the excess energy will be

immediately (usually within 10-14 seconds) emitted in the form of prompt gamma ray. The newly

formed nucleus is often still unstable (radioactive), so it will beta decay to a stable state by emitting

a beta particle and one or more subsequent (delayed) gamma rays with fixed half-life times (Figure

3.2.1). The time frame for the emission of delayed gamma rays ranges from seconds to days, or

even up to months.

Figure 3.2.1. Schematic representation of radioactive capture reaction [17].

Measuring the prompt gamma rays is often experimentally complex for the neutron

activation analysis. However, the delayed gamma rays can be measured relatively easily by HPGe

detectors and later analyzed for individual radionuclide identification. The delayed gamma rays are

extremely important for NAA, as they carry specific decay energy information about the element

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in the material that can be used as a fingerprint to identify this element, including its multiple

isotopes [16]. Radiative capture is denoted by the notation (n,γ).

3.3 NAA Methods

NAA is a powerful, precise, and versatile analytical technique suitable for the analysis of

many types of samples, hence it is actively employed in a wide range of disciplines such as

archaeology, geochemistry, health and human nutrition, semiconductor technology, and

environmental monitoring [18]. Depending on applications, tested samples, experiment conditions

and objectives, NAA can be customized and performed with a different methodology. All NAA

methods use the same principle that was discussed in Section 3.1; however, they differ from one

another in sample irradiation and data analysis. Each method has advantages and disadvantages.

The most common three NAA methods will be briefly discussed in the following sections.

3.3.1 The Absolute NAA Method

The absolute NAA method, also known as instrumental neutron activation analysis

(INAA), determines the absolute elemental or isotopic concentration in the test sample material,

directly using the measured experimental parameters, such as the activity of the irradiated sample

and local neutron flux (Equation 3.3.1.1).

A = N(1 – e-λt)[ σthΦth + σresΦres] (3.3.1.1)

In equation 3.1, A (measured activity), N (number of atoms), λ (decay constant) values are

associated with an irradiated element in the sample, t is the decay time between the end of

irradiation and the beginning of radiation counting, σth (thermal) and σres (resonance) are the neutron

absorption cross-sections, and Φth (thermal) and Φres (resonance) are neutron flux magnitudes

measured at the sample irradiation location [19].

14

This method is very sensitive to the accuracy of measured values, so it is extremely

important to use proper procedure and equipment to obtain adequate results. Thus, for the sample

activity measurement high resolution HPGe detectors are typically used. Nevertheless, the

efficiency and calibration of these detectors might slightly vary, due to the small parameter changes

regarding sample geometry, orientation, etc. One of the disadvantages of INAA is the fact that it is

challenging to avoid small changes during the radiation counting and irradiation, which might

significantly affect the final results. Another challenge within this method is related to measuring

accurate local neutron flux values and determining the exact number of activated atoms due to the

neutron flux exposure. For high precision, it is also important to take into account the self-shielding

of the sample. In practice, it is difficult to maintain identical efficiency and calibration of the

detector and to measure the neutron flux value for each sample location; therefore, this method is

more applicable when there are only a few samples with the same size and relatively simple

elemental composition [20].

The absolute NAA method was not found suitable for this work, since the objective of this

work was a characterization of 22 fracking soil and water samples that have very complex elemental

compositions.

3.3.2 Comparative NAA

Comparative NAA (CNAA), also called the relative calibration method, is another

approach that avoids some of the drawbacks of the absolute NAA method while determining the

concentration of an element/isotope in a sample. In order to perform CNAA, it is necessary to have

rough knowledge regarding the elemental/isotopic content of the test sample and one or more

standard materials of similar content. The sample(s) and standard(s) must be irradiated and counted

under the same conditions. The elemental/isotopic concentrations in the unknown (tested) sample

are determined using Equation 3.3.2.1.

15

𝑤𝑢 = 𝑚𝑠𝐴𝑢𝐷𝑢𝐶𝑢Φ𝑠

𝑚𝑢𝐴𝑠𝐷𝑠𝐶𝑠Φ𝑢 (3.3.2.1)

Where the subscripts u and s refer to the unknown and standard used in the comparison, 𝑤

is the concentration of the element of interest, m is the mass of the sample, A is the measured

activity of the target isotope (including the detector efficiency, the saturation factor, etc.), D is the

decay correction factor, C is the counting correction factor, and Φ is the exposed neutron flux. The

decay correction factor for each sample can be calculated via Equation 3.3.2.2.

𝐷 = exp (−λ𝑡𝑑) (3.3.2.2)

Where λ is the decay constant of the isotope of interest, and 𝑡𝑑 represents the time between

the end of irradiation and beginning of radioactive counting. The counting correction factor is more

important during long radioactive counting, since it is accounting for decay during the measurement

(Equation 3.3.2.3).

𝐶 = 1−exp (λ𝑡𝑚)

λ𝑡𝑚 (3.3.2.3)

The variable 𝑡𝑚is the radioactive counting time.

CNAA has a clear advantage over other NAA methods, if the suitable comparator standard

and the test sample have a similar geometry, background matrix, and trace element composition

[19] [20]. By irradiating test and standard samples together, it is possible to eliminate the necessity

for an accurate knowledge of the neutron flux, assuming that both samples exposed an equivalent

neutron flux. Moreover, using CNAA method, there is no need to evaluate the detector efficiency,

calibration, and counting geometry effects as it was required by INAA. Therefore, the

ascertainment of the composition of trace elements is performed more simply with CNAA

compared to INAA. The main disadvantage of CNAA is that it becomes inapplicable and useless

with the absence of a suitable standard material for the desired sample matrix, which includes the

elements of interest [19].

16

3.3.3 Single Comparator NAA

Single Comparator NAA, also known as k-factor method, is another comparative approach

to perform multi-element analysis of an examined sample. This method was first critically

evaluated by F. De Corte [21] and later was found very useful in the cases where the implementation

of CNAA method is impossible due to the unavailability or exorbitant cost of suitable standard

materials. As the term single comparator refers, this method differs from traditional CNAA by

irradiating, measuring, and comparing only a single element material (mostly gold) as a standard.

In practice, a typical comparator material is a small piece of gold foil or wire. Due to the small size

of the comparator, it can be placed next to the sample during irradiation, thereby ensuring the

equivalent effect of the neutron flux expose (Φ/ Φ*=1) [21].

The determination of the elemental concentration of test sample is based on the ratio of

proportionality factors of the target and comparator elements that defined as k-factor value. The

experimentally-determined k factor can be calculated using the following equations [21]:

𝑘 =𝐴𝑠𝑝

𝐴𝑠𝑝∗ =

𝑀∗

𝑀∙

𝛾

𝛾∗ ∙𝜖𝑝

𝜖𝑝∗ ∙

𝛩

𝛩∗ ∙Φ

Φ∗ ∙𝜎

𝜎∗ (3.3.1.1)

with 𝐴𝑠𝑝 =𝐴𝑝

𝑆∙𝐷∙𝐶∙𝑤 (3.3.1.2)

where <*> sign refers to the single comparator or monitor, and absence of any sign indicated the

unknown sample. The variable Asp is the specific count rate, M is the atomic weight of the irradiated

element, Θ is the isotopic abundance of the target nuclide, γ is the absolute abundance of the

measured γ-ray, ϵp is the full-energy peak efficiency of the detector for the measured γ-ray energy,

Φ is the conventional reactor neutron flux [neutron/(cm2∙sec)], σ is the effective reactor neutron

cross-section [b], Ap is the measured average intensity of the full-energy peak [counts/sec], w is the

weight of the element [g], D is the decay factor, and C is a measurement correction factor. The

activity saturation factor (S) dependent on the decay constant (λ), and the irradiation time, (tirr). The

dependency shown in Equation 3.3.1.3.

17

𝑆 = 1 − exp (−λ ∙ tirr) (3.3.1.3)

Calculations using these experimentally determined k-factors are usually more accurate

than calculations of the absolute calibration method, based on literature data [21]. Nevertheless, the

k-factors are very dependent on the measured experimental conditions, so small variations in

neutron flux rate, detector efficiency, or counting geometry can cause a significant error and make

the method invalid [19]. The k-factor determination requires very precise and laborious

experimental work, once they are available there is no need for preparation of standards for further

analysis. The determined k-factors are assumed to be constant as long as there are no variations in

the quantities given in Equation 3.3.1.1 [21].

Comparing these three NAA methods, it was decided to use the comparative method

(CNAA) in this work for several reasons. First, two suitable soil and sediment standard materials

were available for comparison to the fracking samples. Second, the samples were in forms of

powder (soil and sediment) and crystal (dried water), which makes them easy to encapsulate and

shape to similar geometries. Next, CNAA method is more accurate due to less sensitivity to the

changes in the measurement parameters. Finally, data analysis using the CNAA method was found

relatively more straightforward than other methods.

3.4 Applicability of NAA

Like any other method for determining trace elements, the neutron activation analysis is

not completely universal and applicable to all types of materials. The chemical properties, physical

forms, and physical characteristics of the sample are important accounting factors for applicability

of NAA to a specific work. Another essential factor is the element of interest and nuclear properties

of its isotopes, since it influences to the activation rate (absorption cross-section) and decay

characteristics of produced radionuclide (half-life, gamma ray abundance and energies). Therefore,

the very low Z elements (like H, He, B, Be, C, N, and O) and a few other elements (Tl and Bi) are

18

not suitable for NAA characterizations. Some elements such as lead (Pb) can be determined with

low sensitivity (order of milligrams) and useless for many applications. The sample matrixes with

both high density (high atomic number) and very high neutron absorbing properties are also

unwanted, due to strong gamma-ray self-attenuation. High concentrations of the elements (B, Li,

and U) that emit charged particles (α-radiation) after neutron absorption are also not preferable,

since they cause active thermal heating during the continuous irradiation [15].

Despite all these specifics, neutron activation analysis is widely implemented in numerous

disciplines, such as archaeology (metal, stone, pottery artifacts, etc.), biomedicine (tissue, blood,

venom, etc.), environmental science and related fields (aerosols, fossil fuels and fuel, sediments,

etc.), forensics (bomb debris, bullet lead, shotgun pellet, etc.), geology geochemistry (coal and oil

shale components, cosmo-chemical samples, diamonds, etc.), industrial products (alloys, electronic

material, high purity and high-tech materials, semiconductors, etc.), and nutrition (composite diets,

spices, milk and milk formulae, etc.) [15].

In this work, neutron activation analysis technique is used to characterize the soil and water

samples collected from hydraulic fracturing wellbores. The accurate multi-element analyses of

wastewater and soil samples are extremely important for the regulatory monitoring of oil and gas

(O&G) development. The NAA might be an excellent tool for this analysis because it provides very

accurate qualitative and quantitative results in terms of concentrations of constituent elements,

which are used to ensure that everything is within regulatory limits.

19

Chapter 4

Radiation Science and Engineering Center NAA Facility

The experimental part of this work was conducted on the base of Radiation Science and

Engineering Center (RSEC), which houses multiple nuclear research and education facilities, such

as the Penn State Breazeale Reactor (PSBR), Co-60 Gamma Ray Irradiation Facility, Hot Cells,

Radiochemistry Teaching and Research Laboratory, Subcritical Graphite Reactor Facility, the

neutron beam laboratory, Radionuclear Applications Laboratory (RAL), and Nuclear Security

Education Laboratory[22]. The main missions of the RSEC are education (faculty, staff, students,

and public), training (NRC certified reactor operators, inters), and research (NAA, reactor control,

and various neutron beam techniques, etc.). Moreover, using the Penn State Breazeale Reactor

(PSBR) and other irradiation facilities, the RSEC is also provided irradiation services to other

universities, government entities, and the industry [23].

4.1 The Penn State Breazeale Nuclear Reactor (PSBR)

The Penn State Breazeale Nuclear Reactor (PSBR) is the first licensed and longest

continuously operating university research nuclear reactor in the United States, which reached its

first criticality in 1955. The reactor originally was designed as a material test reactor (MTR) that

uses a plate type fuel made of highly enriched uranium. However, after receiving a license

amendment in 1965, the reactor was converted to a TRIGA (Training, Research, Isotopes, General

Atomics) design that operates on a low-enriched uranium-based pin type fuel [22]. The current

TRIGA MARK III reactor core resides on the bottom of 24 feet deep open-pool that is filled with

approximately 71,000 gallons of deionized water. The core is attached to a bridge on rails, so it can

be moved in several directions throughout the pool, providing the flux flexibility for the out-core

irradiation fixtures. In steady state operation, the maximum power of PSBR is rated as 1 MW and

https://www.rsec.psu.edu/Subcritical_Graphite_Reactor_Facility.aspx

20

the maximum thermal neutron flux in the central thimble is about 3*1013 n/cm2s. The PSBR also

has a pulsing ability at which the reactor power can reach 2000 MW for 10 milliseconds and the

thermal neutron flux peak reaches up to 1016 n/cm2s [19].

In this work, the PSBR core loading number 58 was used for the flux profile determination

(section 4.4) and fracking sample irradiation (section 6.3). The map of the core loading number 58

is shown in Figure 4.1.1.

Figure 4.1.1. A map of the PSBR Core 58 Loading

Figure 4.1.1 shown the pattern of the TRIGA pin fuels that contain 8.5 wt% and 12 wt%

low enriched uranium. Moreover, there are demonstrated the locations of the control rods (F9, H6,

H12, and J9), dry tubes (E6 and E11), and central thimble (H9). The multiplicity of irradiation

21

locations and the flexibility of neutron flux makes PSBR a perfect source for NAA applications

[19].

4.2 RSEC Radionuclear Applications Laboratory

The RSEC also houses radionuclear applications laboratory (RAL) that provides technical

support to radionuclear technique users, such as research personnel and industrial users. The

laboratory has a convenient workstation with all necessary equipment for sample preparation and

post-irradiation handling, four complete high-purity germanium detector (HPGe) systems, two

automatic sample handling systems (ASHS), a pneumatic tube transport (rabbit) system terminus,

a Compton Suppression System and multiple Geiger-Mueller (GM) “Pancake” detectors and

Genie-2000 software installed computers [22] [19]. In the following sections will be discussed the

ASHS and HPGe detector system, which were used for radiation counting in this study.

4.2.1 The Automatic Sample Handling System (ASHS)

The automatic sample handling system (ASHS) is an essential tool for all NAA sample

measurement since it operates in conjunction with a radiation counting system and automates the

process. The automation of radiation counting provides additional reliability and consistency of

measurements because it minimizes the error due to counting statistics by reducing the decay time

between sample measurements.

22

Figure 4.2.1.1. The rotary sample holder of the ASHS (capacity is over 90 samples).

The ASHS has a rotary sample tray that is capable to hold over 90 samples (Figure 4.2.1.1).

The sample tray moves each sample one at a time under a pneumatic lever, which picks up and

places the sample into a nest 2.5 cm above of the HPGe detector. When the counting is completed,

the ASHS automatically picks up the remaining sample and replaces it with a next sample. The

user can select counting parameters, such as counting time (from seconds to days) and the number

of samples to be counted (from 1 to 90).

4.2.2 The Counting System

Once the ASHS moves the sample into sample nest, it is measured with a counting system

that includes a Canberra GC1518 HPGe detector, a digital spectrum analyzer (DSA-2000), and a

23

personal computer (PC) with Canberra’s Genie-2000 software. All these radiation detection and

measurement instrumentations and ASHS are connected as it is demonstrated in Figure 4.2.2.1.

Figure 4.2.2.1. Component diagram of instrumentation layout to perform automated

radiation counting [22].

To reduce the background radiation, the HPGe detector is placed into lead shielding cave

with an internal copper and tin liner. These liners contribute to eliminating X-rays caused by the

interaction of gamma rays with lead. The manufacturer specifications and dimensions of the HPGe

detector shown in Appendix A.

The DSA-2000 is a fully integrated system for high quality spectrum acquisition. It

combines the digital signal processor (DSP), high voltage (HV) power supply, digital stabilizer,

multi-channel analyzer (MCA) memory, and an Ethernet network interface. The DSA-2000 and

the PC are connected via cable, and they communicate through Genie-2000 software, which

visualizes the count information for each channel in the real time.

24

The radiation counting data (gamma spectrums) from each sample was recorded in the PC

and later analyzed using Genie-2000 software from Canberra. Each spectrum was consistently

analyzed, following a fixed sequence that includes eight steps listed in table 4.2.2.1.

Table 4.2.2.1. The outline of the spectrum analysis sequence steps and their purposes [20]

[24].

Step Analysis Type Purpose

1 Peak Locate → 2nd

Identified Difference

To identify peaks that have a statistical significance and

located within a user-defined tolerance value (3.00 keV).

2 Peak Area → Sum/

Non-Linear LSQ Fit

To determine the area under of each peak (sum of counts)

using a non-linear least square fit.

3 Area Correction →

Std. Bkg. Subtract

To correct the area under of each identified peak by subtracting

an estimated background contribution (a standard background

file).

4 Efficiency Correction →

Standard

To correct the area of each identified peak, accounting for the

detector efficiency.

5

Nuclide Identification →

NID with Interference

Correction

To identify nuclides by analyzed energy peaks using the

nuclide library of the software. This algorithm also considers

the interference from nearby peaks.

6 Parent Daughter

Correction

To correct an iteration of a parent-daughter decay chain.

7 Detection Limits →

Curie MDA

To exclude false identified isotopes by calculating the

Minimum Detectable Activity (MDA) for each identified

isotope.

8 Reporting To create and save a Portable Document Format (PDF) report

file, that contains all necessary data for further analysis.

To obtain adequate results from the analysis sequence steps, the detector must be properly

calibrated, for energy and efficiency. Since the detector is surrounded by heavy lead shielding, the

contribution of environmental background radiation is minimal or even negligible. However, the

radiation emitted by irradiated quartz ampoule is also considered as a background. Thus, in order

25

to estimate and subtract the contribution of background radiation, it is important to irradiate and

measure an empty quartz vial under the same conditions as applied for other samples.

4.3 PSBR Irradiation Fixtures

The PSBR has multiple irradiation fixtures that are located inside (two dry tubes and a

central thimble) and outside (a pneumatic transfer system, 2” x 6” irradiation fixture, neutron beam

ports) of the reactor core. Each irradiation fixture is unique in terms of available neutron flux

(magnitude, energy group), geometry (size and shape), location (in respect to the reactor core), and

other irradiation conditions (wet, dry, etc.). Because of this variety, the irradiation fixture can be

selected according to the specific requirements of the experiment. In this work were utilized two

irradiation fixtures: dry tube number one (DT1) and the pneumatic transfer system (PTS).

The dry tubes are cylindrical air-filled tubes that are permanently installed in the reactor

grid plate spacer. To ensure reliable placement in the grid plate, the bottom of dry tubes is made in

an identical way as the fuel pin bottoms. Moreover, to maintain the geometric uniformity of the

reactor core, dry tubes have the same diameter as the fuel pins. The top part of the dry tubes is bent

in a large radius and attached to the reactor pool bridge. The bend is designed to prevent direct

shine of gamma rays to the upper of the pool and to ensure safety during sample loading and

unloading. (Figure 4.3.1) [23] [24]. Naturally, air molecules contain Ar-40, which might turn to

radioactive Ar-41 under neutron exposure. Thus, to isolate the dry tube air from the air in the

facility, the top of the dry tube is enclosed with a rubber plug. The rubber plug can be safely

removed when Ar-41 isotopes, created in the dry tube, completely decays away. It typically takes

six half-life periods of Ar-41 (109.6 minutes) and lasts for about 11 hours [23]. More detailed

information about dry tubes can be found in Danielle K. Hauck's PhD dissertation [24].

26

Figure 4.3.1 Shape and design of the dry tubes (DT) [25].

The pneumatic tube transport system (PTTS), also called the “rabbit” system, is designed

to quickly transfer samples from RAL to the reactor pool for irradiation. In addition, the rabbit

system can rapidly bring the irradiated samples back to the RAL, which allows the analysis of

short-lived isotopes in the sample before their decay. The operation principle of PTTS is based on

compressed CO2, which pushes a sample placed in a standard-sized plastic capsule through a

pneumatic tube. One side of this tube is permanently installed in the reactor pool (near the D2O

tank), and the other end is connected to the stationary terminus in the RAL room (Figure 4.3.2).

27

Due to the influence of strong moderators such as water medium and D2O tank, the samples

irradiated in the PTTS mostly experience a thermal neutron flux. The magnitude of the neutron

fluence rate can be varied by the reactor power alteration and the reactor core movement [23].

Figure 4.3.2 The terminus located in the Radionuclear Applications Laboratory

(RAL).

The maximum irradiation time using PTTS is only 10 minutes.

4.4 Neutron Flux Characterization of the Dry Tube 1 for Core 58 Loading

To use the CNAA method it is sufficient to know that the standard reference material and

the examined samples undergo an identical neutron fluence. If one could assume that the neutron

flux level in the dry tube is uniformly distributed, there would be no need for the neutron flux

characterization since all vials are placed relatively close to each other. However, this is not the

case, and in order to ensure safety and high efficiency with respect to the dose level and the

28

irradiation time, it is necessary to have accurate data on the magnitude of the neutron flux at

different axial levels along the Dry Tube. The entire experiment was initiated in March of 2018

when the PSBR Core 57 was upgraded to Core 58; thus, the previous neutron flux characterization

was invalid, and a neutron flux measurement with the renewed core pattern had to be implemented.

The flux measurement was conducted mostly following the procedure described in the Master

Thesis of Sarah Sarnoski [23]

The flux of thermal neutrons with energy of 0.0253 eV and resonance (epithermal)

neutrons with energies above 0.5 eV within Dry Tube 1 was measured using an aluminum-gold

wire and 1 mm thick cadmium tubing. These materials were selected due to their specific natural

characteristics. First, the aluminum-gold wire contains 0.112% gold, which absorbs a neutron

through the 197Au(n,γ)198Au reaction. The cross-section values for this reaction ([98.65 ± 0.09 barns]

for thermal neutrons and [1550 ± 28 barns] for resonance neutrons) are high enough that it does not

require a lengthy irradiation [26] In fact, the cross-sections are so large that a high concentration of

gold would produce too much activity. Moreover, the main activation products, such as Al-28, Mg-

27 and Na-24, have half-life periods on the order of minutes to hours, which renders them as short-

lived isotopes compared to the 2.7-day half-life of Au-198. This fact can be used to reduce the

impact of the aluminum alloy on the gamma ray spectrum by extending the time between irradiation

and counting for several days. Finally, the activated isotope of gold (Au-198) returns to a stable

state by emitting a 411.8 keV gamma ray with 98.99% intensity [26].Due to its very large thermal

neutron capture cross section (20615 ± 400 barns) the cadmium tubing was employed as a filter to

determine the resonance neutron flux. [26] Sections of the aluminum-gold wire were sealed in the

cadmium tubing so that only neutrons of epithermal energies are absorbed by the aluminum-gold

wires. The entire procedure of the neutron flux characterization can be described in three main

steps: 1) Preparation of the samples and documentations; 2) Irradiation and counting of the wire

samples; and 3) Analysis of the collected data.

29

4.4.1 Preparation of the samples and documentation

The sample preparation for irradiation commenced with measuring and cutting two 20-inch

and two 3-inch pieces of an aluminum-gold wire. Next, the wires and two aluminum holders for

loading were cleaned with ethanol since the irrelevant elements on the surface might be activated

and contribute to the gamma ray spectrum as trace elements. After cleaning, each item was weighed

using an AE Adams AEA-100SG balance. All these and further actions should be performed using

laboratory gloves, because with direct interaction, human skin can contaminate samples with

sodium and other activatable isotopes. Then, the longest piece was stretched and taped to an

aluminum holder which holds the wire steady and straight. (Note: one end of the tape must be

folded to facilitate its removal from the "hot" wire after irradiation.) The second wire was severed

into six half-inch pieces and placed into 1 mm thick cadmium tubing, which was sealed on both

ends. Further, each of the cadmium covered wires were taped to another aluminum holder following

the pattern shown in Figure 4.4.1.1.

The aluminum holders have marks on the surface with a 1-inch scale, which are useful for

tracking the pattern during the taping. (Note: the cadmium tubing was cut off slightly longer than

half-inch wire to facilitate removal of the wire; thus, the cadmium cover should be taped slightly

below the mark since the alignment occurs over the inner wire and not its cover.) Both bare wire

(BW) and cadmium covered wire (CCW) samples where labeled with DT1 (Dry Tube 1) markings.

The irradiation procedure of the PSBR requires an experiment evaluation and authorization

document called Standard Operating Procedure (SOP-5). This document must include the

experimental description, irradiation conditions and the post-irradiation activity prediction

magnitudes. The latter was performed using an activity prediction application, developed by Dr.

Dağistan Şahin, which will be discussed in detail later in this thesis [25]. Using the information

regarding the type and mass of tested materials, irradiation time, and approximate neutron fluence,

the application estimated the total activity and gamma ray exposure rate with respect to distance

30

and time. Based on the obtained data, was determined the appropriate irradiation time, the decay

time, and the counting time. Finally, the SOP-5 was reviewed and approved by PSBR personnel.

Figure 4.4.1.1. The cadmium covered wire positions within Dry Tube 1 with respect

to fuel rod.

4.4.2 Irradiation and Counting of the Wire Samples

Both sample irradiations were scheduled on Monday mornings when reactor starts after

2.5-day weekend breaks. After that amount of time, the xenon neutron poison (Xe-135)

concentration drastically drops down (half-life is 9.14 hours) making the reactor fresh and ready

for clean critical condition. Under this condition the poison contribution to the neutron fluence will

be minimal, and all control rods will be in near-identical positions. Following the schedule, the first

sample irradiation was initiated on March 19, 2018 at 09:44 AM with a bare wire at the reactor

31

power 800 kW. After 7 minutes of nonstop irradiation, the sample was pulled up to six feet above

the core and remained in Dry Tube 1 to decay for 3 days. Then, it was relocated to the shadow

shield corner and checked by Environmental Health and Safety (EHS) personnel. After getting

EHS authorization to work, the sample was placed into a lead tube container and moved to the

Radionuclear Application Laboratory, which is located on the lower floor of the same building.

(Note: the movement took place utilizing a cart to maintain a safe distance between the human body

and the radioactive sample.) The cadmium covered wire sample was irradiated on the next Monday

at 11:56 AM with the same reactor power and irradiation time. The second sample was also

transported to the laboratory following the identical timing and safety procedures.

In order to determine the actual neutron fluence shape within the Dry Tube 1, the samples

had to be prepared for counting. The objective was to cut the 20-inch activated bare wire into 40

pieces of a half-inch and place each piece into small plastic vial. The objective was accomplished

by pulling the sample out from the container and each time cutting one inch of activated wire by

wire cutter. (Note: after each cutting, the sample should be descended back into container to reduce

radiation exposure. Also, an exposure time can be reduced by using one-inch marks on the

aluminum holder as a measuring tape.) Further, using a long tweezer and wire cutter, each one-inch

piece was cut into two identical pieces in a plexiglass shield box that shield the beta particles

emitted by the wire. Finally, all half-inch pieces were placed into plastic vials and numerated from

1 to 40. Since cutting was started from the top of the wire, number 40 corresponds to the bottom of

the dry tube with respect to the reactor core. The cadmium covered wire was processed utilizing

the same method except for the additional step of removing the aluminum-gold wire from the thin

cadmium tubes. Using pliers, the soft cadmium tubes were squeezed and unsealed which allowed

the wire piece to slip out. Cadmium tubes were later disposed in a container of mixed radioactive

and hazardous waste.

32

The sample counting used an automatic sample changer at the Radiation Science and

Engineering Center, which automates the radioactive counting process, saving working hours, and

providing very consistent measurement times. Each enumerated vial loaded in the sample changer

was automatically picked up and placed into a lead cave with an HPGe detector on the bottom. For

each sample, the live counting time was set to 15 minutes, after which the sample was automatically

retrieved from the cave and placed back in the sample changer wheel. Data acquired from each

sample was automatically saved during the counting process. For both counting sessions, the

detector calibration was not performed since in the Radionuclear Application Laboratory the

detectors are calibrated annually, and during the data acquisition periods the last calibration was

still valid.

4.4.3 Analysis of the Collected Data

Ten days after the irradiation, the vast majority of activated isotopes in the samples

decayed, which made the samples less radioactive. Once the wire pieces were safe to handle, each

wire sample was weighed using an AE Adams AEA-100SG balance. The obtained weight values

were used to determine the number of gold atoms in each sample using Equation (4.4.3.1).

𝑁𝐴𝑢 = 𝑤𝑒𝑖𝑔ℎ𝑡 𝑜𝑓 𝑠𝑎𝑚𝑝𝑙𝑒 ∗𝑊𝐴𝑢 ∗ 𝑁𝐴

𝑀𝐴𝑢 (4.4.3.1)

Where WAu is the gold content in the wire, NA is Avogadro’s number, and MAu is molecular

mass of gold. Due to the isotopic abundance of gold, was assumed that initially 100% of gold

contained Au-197. Considering the relatively low neutron flux magnitude and short irradiation

time, it was also assumed that only a small fraction of gold isotopes was activated, and 100% of

the activated isotopes were Au-198.

In order to calculate the neutron fluence, it is necessary to define the saturation activity. In

theory, the saturation activity can be reached only if the irradiation time is too long in comparison

33

with the half-life of the isotope of interest. Since this is not the case, the saturation activity for each

sample was calculated in a corrected version using Equation (4.4.3.2).

𝐴𝑠𝑎𝑡 = 𝐴𝑚𝑒𝑎𝑠 ∗exp (𝜆 ∙ 𝑡𝑏𝑒𝑡)

(1−exp (−𝜆 ∙ 𝑡𝑖𝑟𝑟𝑎𝑑))∗ 𝜀 (4.4.3.2)

Where 𝐴𝑚𝑒𝑎𝑠 is the measured activity of the wire in counts per second, 𝜆 is the decay

constant of the isotope, 𝑡𝑏𝑒𝑡 is the time between irradiation and counting, 𝑡𝑖𝑟𝑟𝑎𝑑 is the irradiation

time, and 𝜀 is the detector efficiency.

Another valuable variable for calculating the neutron flux is the cadmium ratio (CR), which

is defined as the ratio between the saturation activities of bare and cadmium covered wires located

at the same axial level (Equation 4.4.3.3). The neutron flux within DT1 can be calculated using the

fact that the pieces of BW were irradiated with both thermal and resonance (epithermal) neutrons,

while the CCW pieces were activated by only resonance neutrons.

𝐶𝑅 = 𝐴𝑠𝑎𝑡. 𝐵𝑊

𝐴𝑠𝑎𝑡. 𝐶𝐶𝐷=

𝑡ℎ𝑒𝑟𝑚𝑎𝑙 𝑎𝑐𝑡𝑖𝑣𝑖𝑡𝑦 + 𝑟𝑒𝑠𝑜𝑛𝑎𝑛𝑐𝑒 𝑎𝑐𝑡𝑖𝑣𝑖𝑡𝑦

𝑟𝑒𝑠𝑜𝑛𝑎𝑛𝑐𝑒 𝑎𝑐𝑡𝑖𝑣𝑖𝑡𝑦=

𝐴𝑡ℎ + 𝐴𝑠𝑎𝑡. 𝐶𝐶𝐷

𝐴𝑠𝑎𝑡. 𝐶𝐶𝐷 (4.4.3.3)

Using Equation (4.4.3.3), the CR value was calculated at 5.25”, 8.25”, 11.75”, 13.75”,

16.75”, and 20.25” distances from the DT1 bottom.

To determine the thermal neutron flux within the DT1, it is necessary to define the activity

caused by thermal neutrons. The definition of the thermal activity can be derived from the Equation

(4.4.3.3).

𝐴𝑡ℎ = 𝐴𝑠𝑎𝑡. 𝐶𝐶𝐷 ∗ (𝐶𝑅 − 1) = 𝐴𝑠𝑎𝑡. 𝐵𝑊 ∗ (1 −1

𝐶𝑅) = 𝐴𝑠𝑎𝑡. 𝐵𝑊 ∗ 𝐴𝑅 (4.4.3.4)

Where AR (Activity Ratio) is another consequential variable, which was applied to

facilitate the calculations. Using previously determined CR magnitudes, the AR values were also

computed for six axial positions and plotted in Figure 4.4.3.1. Using a least squares method, the

obtained data was fitted to a first-order polynomial equation (see Figure 4.4.3.1), which further was

used to calculate intermediate AR values for all 40 axial positions. Then, those AR values were

applied to Equation (4.4.3.4) to determine the thermal activity within DT1. As it was planned

34

earlier, the thermal activity values were used to calculate the thermal neutron flux magnitudes with

respect to distances from the DT1 bottom. The calculations were performed using Equation

(4.4.3.5).

Φ𝑡ℎ =𝐴𝑡ℎ

𝑁𝐴𝑢∗ 𝜎𝑡ℎ (4.4.3.5)

Where 𝜎𝑡ℎ is the thermal microscopic cross section value, which is equal to 98.65 ± 0.09

barns [26]. The measured shape of the thermal neutron flux within DT1 is depicted in Figure

4.4.3.2.

Figure 4.4.3.1. The AR (Activity Ratio) within the DT1 (Dry Tube 1).

y = 0.0039x + 0.4551

R² = 0.1746

0.35

0.4

0.45

0.5

0.55

0.6

0.65

0 5 10 15 20 25

AR

(A

ctiv

ity R

atio

)

Distance from the DT1 Bottim (in)

35

Figure 4.4.3.2. Measured thermal neutron flux within DT1 (Dry Tube 1).

The thermal neutron flux values at each position were used to calculate the true values of

the resonance neutron flux along the DT1. Equation (4.4.3.6) was used for the resonance neutron

flux calculation.

Φ𝑟𝑒𝑠 = (𝐴𝑠𝑎𝑡. 𝐵𝑊

𝑁𝐴𝑢− 𝜎𝑡ℎΦ𝑡ℎ) /𝜎𝑟𝑒𝑠 (4.4.3.6)

Where 𝜎𝑟𝑒𝑠 is the resonance microscopic cross section value, which is equal to 1550 ± 28

barns [26]. In the same way as the thermal neutron flux, the resonance neutron flux was also plotted

as a function of distance from the bottom of the DT1 (see Figure 4.4.3.3).

0.0E+00

2.0E+12

4.0E+12

6.0E+12

8.0E+12

1.0E+13

1.2E+13

0 5 10 15 20 25

Th

erm

al N

eutr

on

Flu

x (

n/c

m^2

*s)

Distance from DT1 Bottom (in)

36

Figure 4.4.3.3. Measured resonance neutron flux within DT1 (Dry Tube 1).

The peak of the thermal neutron flux was at 10.25 inches above the DT1 bottom and

reached a magnitude 1.09x1013 n/cm2s. The maximum resonance flux was detected at the same

distance and had a value 7.6x1011 n/cm2s. The average thermal and resonance neutron fluences over

all calculated values were 6.95x1012 n/cm2s and 4.7x1011 n/cm2s (see Figure 4.4.3.4).

0.0E+00

1.0E+11

2.0E+11

3.0E+11

4.0E+11

5.0E+11

6.0E+11

7.0E+11

8.0E+11

9.0E+11

0 5 10 15 20 25

Res

on

ance

Neu

tro

n F

lux

(n

/cm

^2

*s)

Distance from DT1 Bottom (in)

37

Figure 4.4.3.4. The thermal and resonance neutron flux peak positions within the

DT1 in regard to a PSBR fuel rod.

The error bars demonstrated in Figure 4.4.3.2 and Figure 4.4.3.3 indicate an average of

±5% estimated error. Due to the percentage error, the error magnitudes vary regarding the neutron

flux values. This specific error was chosen based on Hughes’s statement, which claims that the

minimum error for neutron flux determined at a particulate point is ±5% [25] [27]. Moreover, this

error has been used in last neutron fluence measurement experiments executed at PSBR.

38

Chapter 5

Fracking Soil, Sediment and Wastewater Samples

All samples described in this work were obtained from Dr. Nathaniel R. Warner, an

assistant professor of civil and environmental engineering at the Pennsylvania State University.

There were fifteen solid (soil and sediment) and seven liquid (wastewater) samples, placed in

plastic vials of different shapes and properly labeled with the sample name (Figure 5.1). The solid

samples were in the form of differently colored powder or small (1-2 mm) crystals while liquids

were transparent or light-yellow shaded wastewaters.

Figure 5.1. A picture of all tested fracking soil, sediment, and wastewater samples in

their original plastic containers.

Table 5.1 shows a list of characterized fracking samples in this study with a brief

description.

39

Table 5.1 List of the analyzed samples.

# Sample name Sample description

1 HR SaH Treatment Marcellus

2 BO1 Barite

3 BO2 Barite

4 Sample 01 Solid Mixture of raiobarite and sediment for inter-lab comparisons.




8 AMD cycle 2 Solid from mixture of AMD (acid mine drainage) and brine

9 AMD test 5 Solid from mixture of AMD (acid mine drainage) and brine

10 AMD test 6 Solid from mixture of AMD (acid mine drainage) and brine

11 HR Evop. Test 01 Marcellus evaporation treatment test fluid

12 HR Evop. Test 02 Marcellus evaporation treatment test fluid

13 HR, Evop Test 03 Marcellus evaporation treatment test fluid

14 Raw flowhart solid Solids removed from Marcellus waste prior to treatment

15 FS3 Effluent Centralizes waste treatment effluent

16 Marcellus Flowback Marcellus flowback as part of BAMR (basic agency monitoring

report) project

17 Franklin discharge Centralizes waste treatment effluent

18 Sample 01 Liquid Mixture of liquid brines for inter-lab comparison





According to Dr. Warner, the samples were collected from the hydraulic fracturing

wellbores (mostly based on Marcellus Shale and Blacklick Creek), located within Pennsylvania.

The wellbores are properties of various private companies; therefore, due to restrictions on

proprietary information, the names and exact locations of the wellbores and more detailed

information on test samples are considered confidential information. However, in this section the

treatment details of some samples, which are open to the public, will be discussed.

Three oil and gas wastewater samples (Sample 01 Liquid, Sample 02 Liquid, and Sample

03 Liquid), collected from the Appalachian wells in the northeastern United States, were stored in

the high-density polyethylene (HDPE) containers with a capacity of 20 liters. One liter of each

40

sample was filtered using a cellulose acetate filter (0.45 µm) and placed in a refrigerator at 4°C for

further analysis of anions, such as Cl, Br, and SO4. Next, after acidifying with 5% HNO3 (nitric

acid), the wastewaters were filtered for cation and radioactivity analyses, such as Al, As, B, Ba, Ca

Cr, Cu, Fe, K, Li, Mg, Mg, Ni, Pb, Ra, Sr, U, and Zn. Then, the sub aliquots of each sample were

placed in HDPE vials and shipped to the laboratories for further analysis [1] The detailed

description on the other liquid samples are not provided..

Four solid samples were pulverized and sieved until they reached a grain size of ~1.18 mm

and had a similar matrix with commonly analyzed solid samples for environmental studies of oil

and gas production impacts, such as barite sludge from treatment facilities, shale core or cuttings,

and river sediments impacted by O&G. In this study, these solid samples labeled as: Sample 01

Solid (a stream sediment collected from the Blacklick Creek, western Pennsylvania), Sample 02

Solid (an outcrop collected from a Marcellus Shale), Sample 03 Solid (the mixture of a stream

sediment from the Blacklick creek with a radio-barite sludge) and Sample 04 Solid (the identical

mixture as Sample 03 Solid, but with different ratio). Before the samples were packed and shipped

to the laboratories, they were homogenized using a concrete mixing paddle [1].

41

Chapter 6

The Experiment

6.1 Activity Prediction

Before any sample irradiation, a system of pre-irradiation activities is required, which

includes activity prediction. At the PSBR, activity prediction is usually performed using the

Activity Predictor program developed by Dr. Dağistan Şahin, a former RSEC graduate student [28].

This tool calculates the expected activity for every isotope in a sample, using input values for the

irradiation location (i.e. the neutron flux), irradiation time, sample mass, and sample elemental

content. In addition, this tool can be used to ensure that the post-irradiation gamma ray dose rate

remains within safety limits. Moreover, it can provide a rough image of the expected gamma ray

spectrum with respect to the decay time. The combination of all this knowledge makes it possible

to improve the experiment design regarding high efficiency, radiation safety, and ideal timing.

The tool was developed using the Java programming language. It calculates activities,

exposure rates, and gamma ray spectra of activated samples using both analytical and Monte Carlo

methods [28]. However, to expedite and simplify the calculation process, the full Monte Carlo

algorithm was substituted with a quasi-version, which can accomplish calculation within minutes.

For calculation, this software uses an XML database that includes the databases such as "Tables

for the analysis of neutron activation" (for neutron cross-section data) [29], Berger and Hubbell’s

XCOM photon cross sections database [30], and Lund/LBNL nuclear data search (for isotope decay

data) [31]. The tool predicts the gamma ray spectrum of the irradiated sample using the efficiency

and geometric characteristics of the HPGe detector located at the Radionuclide Applications

Laboratory (RAL) in RSEC. The tool is user friendly since all desired details regarding reactor,

42

experiment, and compound sample properties can be defined on a single graphical user interface

(Figure 6.1.1).

Since most of the examined samples were soil and sediment products, it was decided to use

Buffalo River Sediment and Montana Soil standard reference material (SRM) properties to perform

all post irradiation activity, exposure rate, and spectrum predictions. During that process, the library

of the activity predictor was updated, including a dozen of missing materials such as U, Pb, Hg, N,

H, C, Tl, Ga, Hf, Li, Sr and Th. The most abundant isotopes of these materials were selected and

embedded to the XML database of the software.

Figure 6.1.1. The graphical user interface (GUI) of the Activity Prediction Tool [28].

Finally, the Activity Predictor tool was used to determine the most suitable reactor power,

irradiation, decay, and counting time magnitudes for this experiment.

6.2 The Sample Preparation

The sample preparation procedure of this research was designed by reviewing and updating

the procedure followed in Chad B. Durrant’s Master Thesis [20]. To avoid internal contamination

of the irradiation fixtures in the PSBR, all samples are usually irradiated using the double

encapsulation method, which provides extra safety. Another concern was to choose appropriate

materials for these encapsulations. For brief irradiations (on order of minutes), the samples can be

placed in polyethylene vials; however, a 10-hour irradiation with a reactor power of 800 kW

43

requires a more physically robust and radiation-resistant material since this radiation dose might

lead to unacceptable embrittlement of polyethylene materials [20]. Moreover, the encapsulation

material should have very a small overall neutron capture cross section to reduce the level of

neutron activation. This restriction ensures low dose exposure of personnel during sample handling

and avoids the need for removal of the sample for gamma ray counting. Reviewing the previous

NAA experiments performed at the PSBR, it was decided to use high purity quartz for the first

encapsulation and aluminum foil for the second encapsulation.

The first encapsulation used very high purity quartz tubes, Heraeus Suprasil 310.

According to Heraeus Quartz America, LLC, the impurity concentration of their product is less

than 0.01 ppm (Table 6.2.1).

Table 6.2.1. Typical trace impurities in Heraeus Suprasil 310 quartz glass [32].

Impurities Suprasil-family (ppm)

Aluminum (Al) ≤ 0.010

Calcium (Ca) ≤ 0.015

Chromium (Cr) ≤ 0.001

Copper (Cu) ≤ 0.003

Iron (Fe) ≤ 0.005

Potassium (K) ≤ 0.010

Lithium (Li) ≤ 0.001

Magnesium (Mg) ≤ 0.005

Sodium (Na) ≤ 0.010

Titanium (Ti) ≤ 0.005

Table 6.2.1 shows that some of the quartz impurity elements might also be present in the

fracking samples. Considering the general detection limits of the NAA method, the expected

concentrations of these elements in the fracking samples are higher than 10 ppm. Thus, due to the

difference of at least three order of magnitude, the presence of the listed elements in the pure quartz

can be neglected. In addition, the pure quartz is composed of silicon dioxide molecules, which

consist of two elements with mostly short-lived isotopes that decay away before radiation counting.

44

The Heraeus Suprasil 310 quartz tubes are manufactured with a length of 1500 mm, 6 mm

of external and 4 mm of internal diameter. Using a quartz cutting tool, the tubes were cut to ~45

mm pieces. Then, in the machine shop of the RSEC, each quartz tube piece was sealed at one end,

using a propane-oxygen torch. After cooling down, the half-sealed ampoules were cleaned with

ethanol and weighed using the AE Adams AEA-100SG balance. The next step was placing the

SRM, fracking soil and water samples into the ampoules. The SRMs and solid fracking samples

were directly placed into ampoules and numerated for sample identification. (Note: To maintain

high accuracy, it is necessary to clean all used equipment with ethanol and change laboratory gloves

after handling each sample. Also, to avoid scorching of the samples during the top sealing, the top

1 cm of the ampoule should be wiped from the internal dust residue.) Then, the ampoules with

sample containment were initially weighed, dried in a laboratory oven for 2 hours at 110 °C, and

weighed again. After weighing, the dried samples were placed into a desiccator until the open end

of the ampoules were sealed using the identical technique. All weighing records are demonstrated

in Table 6.2.2. The “AMD test 5” and “AMD test 6” solid samples were exception from this general

procedure, since they contained too much water (more information on these samples can be found

in Table 5.1). Thus, they remained in the oven for two days at 110 °C temperature.

Liquid samples contain a high concentration of hydrogen, so to obtain accurate results from

the NAA, they must be dehydrated before irradiation. However, the dehydration process might also

influence the original elemental content of samples by driving off volatile elements such as Se, if

the evaporation temperature is too high. Thus, Dr. Amanda Johnsen and her student Colleen

Mulhollan conducted the research, using Se concentration in a sample. For this research an AAS

(atomic absorption spectroscopy) standard solution was purchased with a selenium concentration

of 1000 µg / ml ± 1% and a molecular formula of Se in 5% HNO3 [33]. Using this solution three

test samples were prepared with a selenium concentration of 10 ppm, 100 ppm, and 1000 ppm

(Figure 6.2.1 (on the left)). Next, these samples were placed into small polyethylene vials, sealed

45

on top, and called “original samples,” as shown in Figure 6.2.1 (on the right). Then three more

identical vials were filled with these samples and dehydrated under a heating lamp, which offers a

lower evaporation temperature than might be typically used on a laboratory hot plate. Before the

top-sealing, the vaporized vials were refilled with nitric acid to maintain an equivalent geometry

during irradiation and radiation counting. Next, each sample was properly labeled, placed into

rabbit capsule, and irradiated in the pneumatic tube transport system (PTTS).

Figure 6.2.1. The Se sample with a concentration of 100 ppm (on the left) and a

standard PTTS capsule with the Se sample loading (on the right).

The original and evaporated samples with 10 ppm and 100 ppm selenium concentrations

were irradiated for three minutes. Due to a relatively high concentration of selenium, irradiation of

1000 ppm samples lasted only for a minute. After irradiation, each sample was counted for 60

seconds using the HPGe detector. The saturation activities of selenium for each sample is

demonstrated in Table 6.2.2.

The maximum error detected in this research was 15.52 percent, which is within the

acceptable range. Therefore was concluded that dehydration using the heat lamps would not vastly

affect the initial elemental Se concentrations of the sample.

46

Table 6.2.2. Comparison of selenium saturation activities within original and evaporated

samples.

Concentration Saturation activity (Bq)

Error (%) Original Evaporated

10 ppm 70874.4 59877.5 15.52

100 ppm 690955.5 711527.2 2.98

1000 ppm 6166183.4

6063360.4 1.67

6104804.1 0.99

6107883.1 0.94

Using a measurement pipette, 10 ml of each liquid sample was placed into 20 ml glass vial

and weighed. During the weighing process the fifth digit on the balance screen was continuously

changing every 4-5 seconds, since the balance was sensitive enough to react to evaporation of the

liquid. Consequently, the values shown in Table 6.2.3 might be slightly different from the true mass

values. Next, all glass vials were placed under heating lamp for at least 8 hours until they turned to

solid crystals. After dehydration, the samples were weighted again. The challenge with fifth digit

was faced here too, since the dry crystals were absorbing moisture from the air. Then, the crystals

were powdered and handled as solid samples, following the same procedure. The “Sample 04

liquid” and “Sample 05 liquid” samples have completely evaporated, remaining only a thin layer

of salt on the vial walls. Therefore, those samples were transferred from the vial to the half-ampoule

using distilled water, which later also was evaporated under hitting lamp.

47

Table 6.2.3. The identification numbers and weights of the test samples.

# Sample name Empty ampule (g) Wet mass (mg) Dry mass (mg)

Bucket #1

1 Buffalo River Sediment 1.63443 72.45 72.15


3 Montana Soil (2711) 1.69375 72.76 71.5

4 Montana Soil (2711) 1.61128 71.17 70.3

5 HR SaH 1.6334 72.97 68.93

6 BO1 1.62692 73.16 72.84

7 BO2 1.61896 72.47 71.95

8 Sample 01 Solid 1.60217 72.57 72.06

9 Sample 02 Solid 1.57413 71.33 70.74

10 Sample 03 Solid 1.59116 72.62 72.29

11 Sample 04 Solid 1.61689 72.97 72.14

12 AMD cycle 2 1.62174 73.16 70.13

13 AMD test 5 1.64298 74.66 66

14 AMD test 6 1.67561 74.31 65.3

15 HR Evop. Test 01 1.62833 72.58 72.48

16 HR Evop. Test 02 1.59502 72.08 71.93

Bucket #2



3 Montana Soil (2711) 1.64098 72.58 71.18

4 Montana Soil (2711) 1.6609 73.72 72.11

5 Raw flowhart solid 1.62973 80.48 77.8

6 FS3 Effluent 1.63063 72.73 71.77

7 Marcellus Flowback 1.63067 ------ 94.74

8 Franklin discharge 1.73588 ------ 80.47

9 Sample 01 liquid 1.64267 ------ 84.26

10 Sample 02 liquid 1.66597 ------ 78.79

11 Sample 03 liquid 1.64206 ------ 88.05

12 Sample 04 liquid 1.69295 10.01588 6.93

13 Sample 05 liquid 1.70506 10.0027 6.94

14 HR, Evop. test 03 1.67235 73.62 73.24

In the following Figures 6.2.2 and Figure 6.2.3 all ampullated SRM, soil, sediment and

wastewater samples depicted by bucket and identification numbers.

48

Figure 6.2.2. A picture of fracking and SRM samples placed into Bucket #1.

Figure 6.2.3. A picture of fracking and SRM samples placed into Bucket #2.

49

For the second encapsulation used aluminum foil and an aluminum bucket. The sealed

quartz ampoules were cleaned with ethanol, wrapped in aluminum foil, and labeled with

identification numbers. To load samples in the Dry Tube 1, the aluminum buckets were made by

the PSBR machinist, Brian Bennett (see Figure 6.2.4). All samples were placed into aluminum

buckets following the order depicted in Figure 6.2.4.

Figure 6.2.4. The aluminum bucket dimensions and sample loading patterns.

The empty ampoules placed between samples were later used for background

measurements.

The short irradiation of Bucket #2 content used a plastic bucket of the same size, which

reduces exposure dose and facilitated the sample handling.

6.3 The Sample Irradiation and Counting

With the fracking water and soil samples, Dr. Nathaniel R Warner also provided a list of

23 elements of interest from across the periodic table. Since each NAA-relevant isotope has a

unique neutron capture cross section value, it is important to ensure that the isotopes were

50

influenced by an adequate neutron flux for a sufficient irradiation time. Moreover, the half-lives of

the product radioisotopes vary from a several minutes to hundreds of days. Therefore, to make the

irradiation and counting more efficient, the list of isotopes was split into three groups, such as short-

lived (T1/2 ≤ 15 hours), intermedium-lived (15 hours < T1/2 < 7 days), and long-lived (T1/2 ≥ 7

days) radioisotopes. The short-lived isotopes can be observed only within a short time frame after

irradiation; thus, to reduce exposure dose of personnel and to safely handle the samples, there is no

need for a long irradiation at high reactor power. Moreover, this allows to avoid detector

overwhelming which causes a tremendous dead time. On the other hand, intermediate- and long-

lived radioisotopes require a long irradiation in the order of hours, since they need to be activated

enough to provide a sufficient number of counts on the gamma ray spectra. After a long irradiation,

the samples should be held for several days before counting. After this decay period, all short-lived

isotopes decay away, which is important for both personnel safety and counting statistics. In other

words, a fewer number of radioactive signals causes a lower Compton continuum, which enables

more accurately identify low-energy gamma photopeaks without being obscured. For these reasons,

the entire irradiation process had to be performed in two stages with short and long periods. The

time for both irradiation periods was determined by reviewing the previous similar work [20] and

the results obtained from the activity predictor. Later, all of the activity prediction information was

documented in the SOP-5 (Standard Operating Procedure), which later was reviewed and approved

by PSBR personnel.

On April 18, 2018 at 9:41 AM was performed the first irradiation, that lasted for 6 minutes

at a reactor power of 110 kW. According to calculations based on irradiation time and known

thermal neutron flux in DT1, the total thermal neutron fluence during the entire irradiation reached

the value of ~7.645x1014 n/cm2. After irradiation, the aluminum bucket was pulled up and remained

in DT1 for two hours. Then, it was relocated to the shadow shield corner, checked by

Environmental Health and Safety (EHS) personnel, and remained there for another six hours (due

51

to the high exposure rate). During that 8-hour decay period, near the shadow shield corner was set

up a workplace, which included a Plexiglas beta particle shield and two lead caves (see Figure

6.3.1). To minimize the radiation exposure dose, the samples were placed in one of the lead caves

to be extracted from the aluminum bucket using two large (~30 cm) tweezers. Then, to maintain

the safe distance, the empty bucket was relocated back to the shadow shield corner. Next, each

sample was individually moved to the second cave and unwrapped from the aluminum foil, which

later was disposed in a plastic waste bag behind the lead cave. Corresponding to the identification

number, the samples were placed into numerated polyethylene vials in a sample holder. (Note: a

pancake detector and an ion chamber were used to monitor the exposure rate during the entire

sample handling process.) Finally, all samples were transported to the Radionuclear Application

Laboratory, loaded into the automatic sample changer, and counted with an HPGe detector for 15

minutes. Due to a high concentration of Na-24, some samples demonstrated high deadtimes (tdead

≥ 10%). Thus, those samples and corresponding the SRM samples were counted over and over

until reasonable results were obtained. After counting all samples were wrapped, labeled, and

loaded into the aluminum bucket again for the next irradiation.

Figure 6.3.1. The workplace that was set up near the shadow shield corner. It was

used to prepare irradiated samples for gamma ray counting.

52

The long irradiation was triggered on April 24, 2018 and ended at 4:58 PM of the next day,

taking two reactor operation periods. The irradiation was executed at a reactor power of 800 kW

for 10 hours. The irradiation split for two days would not make any difference, since comparators

and unknowns would experience the same exposure and would be analyzed using the CNAA

method. During the entire irradiation the total thermal neutron fluence magnitude was estimated as

~2.5x1017 n/cm2. After this irradiation the aluminum bucket remained in the DT1 for the five days

until most of the short-lived isotopes decayed. The samples then were prepared for counting,

following the same procedure that was used for the first irradiation. In order to acquire a sufficient

data on gamma-ray signals from intermediate- and long-lived radioisotopes, each sample was

counted for 50 minutes. As predicted, some of the samples demonstrated a very high deadtime

value, so they were recounted after five days with the identical counting time. The next count was

set for 3 hours and was performed when 22 days had elapsed from the day of irradiation. The 3-

hour measurement allowed the collection of an adequate number of counts and facilitated the

identification of the isotope through the gamma-ray spectrum. Due to the high percentage of

deadtime, a 3-hour count was conducted two more times with those particular samples.

Since polyethylene is a sufficiently durable material for short irradiations, it was decided

to replace the aluminum bucket with a polyethylene one with the identical size. This decision allows

to reduce the post-irradiation activity rate and to shorten the decay time between irradiation and

counting. As discussed in section 6.2, most of Bucket #2 content are crystal samples that were

obtained due to dehydration of the liquid samples. Thus, it was assumed that these samples contain

a high concentration of salt elements. Based on this assumption and the information collected from

the first irradiation, it was agreed to decrease the reactor power from 110 kW to 10 kW for the third

irradiation. The third irradiation was started on May 9, 2018 at 8:30 AM and continued for 6

minutes. The total neutron fluence for this irradiation was evaluated as 6.95x1013 n/cm2. After

irradiation, the polyethylene bucket with the samples remained in the DT1 for two hours. Then, the

53

Bucket #2 content was prepared for counting in the same manner as previous samples. Each sample

was counted for 15 minutes, after which all samples were again wrapped in aluminum foil and

loaded into the aluminum bucket.

The last irradiation took place within 10 hours at a reactor power of 800 kW. As in the

second exposure, it also lasted for two days and ended May 11, 2018 at 3:43 PM. The fourth

irradiation samples were prepared and counted using the technique and timing that were used for

the samples of the second irradiation.

54

Chapter 7

Experimental Results

7.1 Data Analysis

After collecting all the output gamma spectra from the experiments described in Chapter

6, the data were analyzed using the comparative neutron activation analysis (CNAA) methodology

(Section 3.3.2). Using the information provided by Dr. Nathaniel R Warner regarding the rough

elementary content of the sample matrices, a list of target elements for this study was compiled.

The list includes such elements as Al, B, Na, K, Mg, Ca, Sr, Ba, Cr, Mn, Fe, Ni, Cu, Zn, As, Np

(uranium activation product), Pb, Pa (thorium activation product), Cl, Br, Se, Ag, and Hg. Then,

the corresponding radionuclides of interest were selected for these elements, taking into account

the natural abundance, neutron absorption cross-section, half-life period, and gamma ray emission

intensities and energies. Due to the multiplicity of required characteristics, it was not possible to

determine NAA suitable isotopes for all these elements. For example, Mg-27, Ag-110, Al-28, Br-

80 (very short half-life period), Mg-28, Np-239 (low gamma ray emission energy), Ni-65 (low

natural abundance), Cu-64, Ag-110m (low absolute γ-ray intensity), Pb-209 (no gamma ray

emission), etc. However, after going through all collected gamma ray spectra, there was noticed

the often presence of europium, lanthanum, rubidium, scandium, and cesium isotopes (Figure

7.1.1); thus, it was decided to expand the list of target elements. All target elements and the

corresponding radionuclides are demonstrated in Table 7.1.1, which also contains the gamma-ray

energies used in this research.

55

Table 7.1.1. The list of elements of interest and their radionuclides with gamma-decay

energies used in this study

Element Nuclide Energy (keV)

Chlorine Cl-39 1267.191

Manganese Mn-56 846.764

Europium Eu-152m 841.63

Potassium K-42 1524.6

Sodium Na-24 1368.626

Arsenic As-76 559.1

Lanthanum La-140 1596.21

Calcium Ca-47 1297.06

Barium Ba-131 496.32

Rubidium Rb-86 1077.1

Protactinium Pa-233 311.9

Chromium Cr-51 320.08

Iron Fe-59 1099.24

Mercury Hg-203 279.19

Strontium Sr-85 514.005

Scandium Sc-46 1120.54

Selenium Se-75 264.65

Zinc Zn-65 1115.54

Cesium Cs-134 604.72

The quantitative analysis for these nuclides was performed by using the CNAA method

and certified Buffalo River Sediment concentration values. After determining the concentrations

of trace elements in each sample, the error analysis was conducted via Equation 7.1.1 [20].

𝜎𝜔 = 𝜔√(𝜎𝐴

𝐴)

2+ (

𝜎𝑚

𝑚)

2+ (

𝜎𝑡𝑑

𝑡𝑑)

2 7.1.1.

Where 𝜎𝜔 is the total error or standard deviation of the weight percent, 𝜔 is the

concentration of the element in the sample, A is the measured decay rate, 𝜎𝐴 is the counting

uncertainty obtained from the Genie 2K report, m is the mass of the sample, 𝜎𝑚 is the weighing

error (0.03 mg for the AE Adams AEA-100SG balance), 𝑡𝑑 is the decay time, and 𝜎𝑡𝑑 is the error

associated with decay time (was taken as 60 seconds, due to the uncertainty at the exact (in order

of seconds) end of irradiation time). All calculated values were arranged in tables for each sample,

as shown in Table 7.1.2.

56

Table 7.1.2. Trace element concentrations of Solid 01 Sample.

Element Nuclide Half-life (h) Short

(wt%) Error

Medium

(wt%) Error

Long

(wt%) Error

Chlorine* Cl-39 0.937 3.47E-02 5.86E-03

Manganese Mn-56 2.579 5.93E-02 3.72E-03

Europium Eu-152m 9.312 5.39E-05 2.59E-06

Potassium K-42 12.355 5.50E-01 2.90E-02

Sodium Na-24 14.997 1.19E-01 1.35E-03

Arsenic As-76 26.24 1.98E-03 1.30E-04

Lanthanum La-140 40.285 1.30E-03 1.65E-04 1.28E-03 6.53E-06 1.12E-03 1.75E-04

Calcium Ca-47 108.86 BDL --- BDL ---

Barium Ba-131 276 1.24E-01 2.05E-03 1.01E-01 1.84E-03

Rubidium Rb-86 447.41 3.58E-03 1.42E-04 2.75E-03 9.60E-05

Protactinium Pa-233 647.4 6.68E-04 6.67E-06 4.84E-04 2.81E-06

Chromium Cr-51 664.9 7.52E-03 1.30E-04 1.46E-02 1.82E-04

Iron Fe-59 1067.9 5.50E+00 2.00E-02 7.51E+00 1.65E-02

Mercury Hg-203 1118.3 BDL --- 7.02E-05 8.18E-06

Strontium Sr-85 1556.4 1.74E-01 4.91E-03 1.18E-01 3.07E-03

Scandium Sc-46 2010.9 6.00E-04 2.28E-06 6.03E-04 1.27E-06

Selenium Se-75 2874.7 4.89E-05 2.19E-05 8.95E-05 6.94E-06

Zinc Zn-65 5854.3 7.70E-03 2.39E-04 7.72E-03 1.23E-04

Cesium* Cs-134 18091.1 1.74E-04 7.84E-06 1.72E-04 2.32E-06

The <*> sign in the first column represents non-certified values that have not passed

quality control; also, the BDL stands for below detection limit. The columns <Short>, <Medium>,

and <Long> contain the weight percent magnitude of the element in the first column after decay

period of 8 hours-2 days, 6-12 days, and 22-28 days respectively. The weight percent (wt%) and

can be easily converted to ppm (μg/g), using Equation 7.1.2.

𝑆𝑎𝑚𝑝𝑙𝑒 𝑚𝑎𝑠𝑠 × 𝑤𝑡% × 106 = 𝑝𝑝𝑚 7.1.2

The tables with trace element compositions of other samples can be found in Appendix D. Using

the values from these tables, the trace element concentrations of all fracking solid and liquid

samples were summarized and arranged in two tables (Table 7.1.3 and Table 7.1.4). In these tables

were selected the values with the lowest quality control error (will be discussed in the next

section). Figure 7.1.1. shows an example gamma ray spectrum with labeled gamma ray

photopeaks. The spectrum obtained from the 50-minute counting of Solid 01 Sample with decay

57

period of 12 days. The data from spectrum was used to fill sixth and seventh columns of Table

7.1.2.

Figure 7.1.1. A gamma spectrum obtained from counting Solid 01 Sample after the

'medium' decay period.

Figure 7.1.2. shows another example gamma ray spectrum with labeled gamma ray

photopeaks. The spectrum obtained from the 50-minute counting of Liquid 01 Sample with decay

period of 12 days. The single and double escape peaks at the energy of 2243 keV and 1732 keV

produced due to high concentration of Na-24 isotope, which emits 2754 keV gamma rays.

58

Figure 7.1.2. A gamma spectrum obtained from counting Liquid 01 Sample after the

'medium' decay period.

59

Table 7.1.3. Trace element concentrations of the fracking samples (Part 1). The values are given in weight percent (wt%).

Element Chlorine* Manganese Europium Potassium Sodium Arsenic Lanthanum Calcium Barium

HR SaH BDL BDL BDL 8.94E-02 3.08E+01 4.93E-02 BDL 1.40E+00 7.53E-03

BO1 2.70E-02 1.99E-02 BDL 1.48E-01 1.24E+00 BDL BDL 1.44E+00 4.45E+01

BO2 3.17E-02 5.68E-03 BDL 2.50E-01 1.01E+00 BDL BDL 1.19E+00 4.53E+01

Sample 01 Solid 3.47E-02 5.93E-02 5.39E-05 5.50E-01 1.19E-01 1.98E-03 1.28E-03 BDL 1.24E-01

Sample 02 Solid 4.88E-02 BDL 1.53E-04 1.42E+00 5.21E-02 2.64E-03 2.54E-03 BDL 5.63E-02

Sample 03 Solid 4.97E-02 4.65E-02 6.15E-05 5.67E-01 8.71E-02 1.66E-03 1.49E-03 BDL 4.23E-01

Sample 04 Solid BDL 3.55E-02 9.60E-05 7.04E-01 1.09E-01 1.65E-03 1.77E-03 BDL 1.86E-01

AMD cycle 2 BDL 4.37E-03 BDL 1.34E-01 4.78E+00 BDL BDL 2.98E+00 3.43E+01

AMD test 5 BDL BDL BDL BDL 1.85E+01 BDL BDL 3.58E+00 6.24E+00

AMD test 6 BDL BDL BDL BDL 1.94E+01 BDL BDL 5.62E+00 9.83E+00

HR Evop. Test 01 BDL 2.50E-02 8.34E-06 2.18E-01 6.50E-02 5.51E-03 3.35E-04 1.35E+00 5.86E+01

HR Evop. Test 02 BDL 4.01E-02 1.44E-05 3.02E-01 9.36E-02 5.35E-03 6.00E-04 1.59E+00 5.32E+01

Raw flowhart solid BDL 7.97E-02 BDL 9.23E-01 1.29E+00 3.93E-03 1.02E-03 3.67E+00 3.33E+00

FS3 Effluent 1.34E-03 1.84E-01 BDL 6.35E-01 1.72E+00 BDL 1.46E-03 7.74E+00 5.87E+00

Marcellus Flowback BDL 2.15E-03 BDL 2.46E-02 1.46E+01 5.79E-04 BDL 3.79E+00 2.15E+00

Franklin discharge BDL BDL BDL BDL 1.23E+01 BDL BDL 8.00E+00 7.70E-03

Sample 01 liquid 1.36E-02 4.06E-03 BDL BDL 1.86E+01 BDL BDL 5.60E+00 4.69E-01

Sample 02 liquid BDL 5.51E-03 BDL BDL 1.66E+01 BDL BDL 5.64E+00 5.12E-01

Sample 03 liquid BDL 1.87E-02 BDL 8.01E-01 1.14E+01 BDL BDL 1.04E+01 2.60E-03

Sample 04 liquid 1.12E+00 2.35E-03 BDL 4.17E-01 2.86E+00 BDL 1.29E-04 1.93E+00 3.25E-03

Sample 05 liquid BDL 1.71E-03 BDL BDL 3.32E+00 BDL 6.19E-06 4.48E+00 4.85E-02

HR, Evop test 03 BDL 7.55E-02 BDL BDL 8.09E-01 BDL 1.34E-06 6.83E-01 4.27E+01

60

Table 7.1.3. Trace element concentrations of the fracking samples (Part 2). The values are given in weight percent (wt%).

Element Rubidium Protactinium Chromium Iron Mercury Strontium Scandium Selenium Zinc Cesium*

HR SaH 1.01E-04 BDL BDL 6.90E-04 BDL 2.37E-01 3.82E-07 BDL 6.76E-05 2.28E-05

BO1 5.64E-04 BDL 1.26E-02 1.43E-02 BDL 9.14E+00 6.33E-07 BDL 4.67E-04 7.69E-06

BO2 3.75E-04 7.59E-06 BDL 1.14E-02 BDL 6.56E+00 4.70E-07 BDL 4.04E-04 BDL

Sample 01 Solid 2.75E-03 6.68E-04 7.52E-03 7.51E+00 7.02E-05 1.74E-01 6.00E-04 8.95E-05 7.72E-03 1.72E-04

Sample 02 Solid 7.45E-03 1.00E-07 6.02E-04 1.16E+00 2.63E-04 0.00E+00 1.59E-03 5.58E-04 BDL BDL



AMD cycle 2 4.08E-04 BDL 1.43E-02 1.16E+00 BDL 6.51E+00 3.35E-06 BDL 9.83E-03 1.65E-05

AMD test 5 BDL BDL BDL 2.11E-01 BDL 7.05E-01 6.82E-06 BDL 1.27E-04 2.40E-05

AMD test 6 4.82E-04 BDL BDL 4.52E-01 BDL 1.39E+00 4.31E-07 BDL 5.01E-04 BDL

HR Evop. Test 01 9.13E-04 BDL BDL 9.52E-01 BDL 7.31E-02 9.84E-05 3.93E-05 1.69E-02 BDL

HR Evop. Test 02 1.46E-03 1.67E-03 BDL 1.19E+00 BDL 3.04E-01 1.57E-04 6.20E-05 5.94E-03 BDL

Raw flowhart solid 4.60E-03 2.60E-04 5.87E-03 5.56E+00 6.73E-05 5.88E-01 4.50E-04 4.08E-04 1.35E-02 1.90E-03

FS3 Effluent 3.76E-03 4.67E-04 6.03E-03 1.61E+00 BDL 1.40E+00 4.71E-04 2.68E-04 7.20E-03 1.77E-04

Marcellus Flowback 3.68E-04 1.94E-05 BDL 8.50E-02 BDL 1.53E+00 1.19E-06 BDL 4.62E-04 1.22E-04

Franklin discharge BDL BDL BDL BDL BDL 1.70E-01 BDL BDL BDL 6.78E-06

Sample 01 liquid 4.63E-04 BDL BDL 4.09E-02 BDL 1.63E+00 BDL BDL BDL 9.76E-05

Sample 02 liquid 5.96E-04 BDL BDL 3.46E-02 BDL 1.71E+00 BDL BDL BDL 8.66E-05

Sample 03 liquid 9.98E-04 BDL BDL 6.00E-02 BDL 2.43E-01 4.05E-08 BDL 6.44E-04 4.30E-05

Sample 04 liquid BDL BDL BDL 2.99E-02 1.26E-04 2.65E-02 3.59E-06 2.00E-04 3.27E-03 BDL

Sample 05 liquid BDL BDL BDL 2.06E-03 BDL 4.99E-01 4.86E-07 BDL BDL 1.15E-05

HR, Evop test 03 5.39E-04 3.17E-03 2.31E-02 1.69E-02 BDL 2.26E+01 2.39E-07 BDL 2.24E-03 BDL

61

7.2 Quality control

All calculated trace element concentration values need to be checked and verified by

quality control analysis. In this study, quality control was conducted using eight standard reference

samples that contain Buffalo River Sediment and Montana Soil materials. These materials were

certified by National Institute of Standards and Technology (NIST), and their certificates of

analysis are attached to Appendix B. The SRMs were chosen due to similarity in trace element

composition with fracking soil, sediment, and wastewater samples. All elements of interest from

Table 7.1.1 were found in both SRMs, except chlorine that is absent in Montana Soil. Thus, the

chlorine concentration values in each sample did not go through the quality control and were

declared as non-certified.

Both standard reference samples were placed in each level of each bucket and irradiated

with other fracking samples. To perform a more accurate quality control analysis, reference samples

at each level were located as close as possible to each other. In this way, SRMs will be exposed by

the identical neutron fluence, avoiding the spatial neutron flux fluctuation during irradiation. After

irradiation, the SRMs were sequentially counted to minimize any errors that might be caused by a

difference in the decay period. All collected data on the Montana soil SRM were then analyzed in

the same way as fracking sample data. The results were compared to the NIST certified values and

shown in Table 7.2.1 and Table 7.2.2.

62

Table 7.2.1. A summary of quality control analysis for Bucket#1.

Element Nuclide

NIST

value

(wt%)

Measured value

Level 1

(wt%) Error (%)

Level 2

(wt%) Error (%)

Chlorine* Cl-39 --- --- --- --- ---

Manganese Mn-56 6.38E-02 6.95E-02 8.86 5.69E-02 10.78

Europium Eu-152m 1.10E-04 9.54E-05 13.27 1.00E-04 8.91

Potassium K-42 2.45E+00 2.33E+00 5.04 2.45E+00 0.12

Sodium Na-24 1.14E+00 1.03E+00 9.90 1.07E+00 6.42

Arsenic As-76 1.05E-02 1.05E-02 0.26 1.07E-02 1.84

Lanthanum La-140 4.00E-03 4.27E-03 6.68 3.45E-03 13.64

Calcium Ca-47 2.88E+00 3.01E+00 4.38 2.73E+00 5.14

Barium Ba-131 7.26E-02 7.32E-02 0.88 7.19E-02 1.01

Rubidium Rb-86 1.10E-02 1.06E-02 13.52 1.11E-02 0.74

Protactinium Pa-233 1.40E-03 1.24E-03 11.12 1.32E-03 5.40

Chromium Cr-51 4.70E-03 4.39E-03 6.66 4.60E-03 2.03

Iron Fe-59 2.89E+00 3.16E+00 9.46 3.27E-02 6.72

Mercury Hg-203 6.25E-04 6.28E-04 0.52 4.80E-04 23.20

Strontium Sr-85 2.45E-02 2.20E-02 10.47 1.80E-02 26.73

Scandium Sc-46 9.00E-04 9.27E-04 3.05 9.00E-04 0.01

Selenium Se-75 1.52E-04 1.48E-04 2.89 1.63E-04 7.30

Zinc Zn-65 3.50E-02 3.27E-02 6.72 3.28E-02 6.47

Cesium* Cs-134 6.10E-04 --- --- --- ---

Table 7.2.2. A summary of quality control analysis for Bucket#2

Element Nuclide NIST value

(wt%)

Measured value

Level 1

(wt%) Error (%)

Level 2

(wt%) Error (%)

Chlorine* Cl-39 --- --- --- --- ---

Manganese Mn-56 6.38E-02 6.97E-02 9.18 6.68E-02 4.63

Europium Eu-152m 1.10E-04 1.03E-04 6.59 1.10E-04 0.43

Potassium K-42 2.45E+00 2.34E+00 4.36 2.35E+00 3.89

Sodium Na-24 1.14E+00 1.12E+00 1.60 1.06E+00 6.61

Arsenic As-76 1.05E-02 1.63E-02 55.16 8.80E-03 16.22

Lanthanum La-140 4.00E-03 3.55E-03 11.14 3.37E-03 15.78

Calcium Ca-47 2.88E+00 2.91E+00 0.87 3.17E+00 10.16

Barium Ba-131 7.26E-02 6.88E-02 5.24 6.82E-02 6.08

Rubidium Rb-86 1.10E-02 1.15E-02 4.32 1.14E-02 4.05

Protactinium Pa-233 1.40E-03 1.40E-03 0.17 1.40E-03 0.16

Chromium Cr-51 4.70E-03 4.81E-03 2.26 4.72E-03 0.52

Iron Fe-59 2.89E+00 3.02E+00 4.56 2.83E+00 2.25

Mercury Hg-203 6.25E-04 2.66E-04 57.36 BDL ---

Strontium Sr-85 2.45E-02 2.13E-02 13.00 1.03E-02 57.88

Scandium Sc-46 9.00E-04 1.00E-03 11.65 9.41E-04 4.50

Selenium Se-75 1.52E-04 BDL 2.04E-04 34.19

Zinc Zn-65 3.50E-02 3.46E-02 1.12 3.43E-02 2.17

Cesium* Cs-134 6.10E-04 --- --- --- ---

63

Cesium is another non-certified element, since it was not determined in any Montana Soil

gamma ray spectra. Some of these concentration values are significantly different from the NIST

specified concentration values. The concentrations of these elements are extremely small (on the

order of ppm); thus, even small deviation in the ppm range can lead to a significant difference in

the results, which are given in weight percent values. In addition, the NIST certified and

experimentally measured concentration values of each element have the same order of magnitude.

7.3 Interlaboratory Comparison of Results

A study on the accuracy of methods for reporting inorganic element concentrations and

radioactivity in oil and gas wastewaters from the Appalachian Basin was conducted at the Civil &

Environmental Engineering Department of the Pennsylvania State University. Eight academic, six

commercial, and one government laboratories throughout the United States, Canada, and Germany

participated in the study. Each laboratory was instructed to characterize three oil and gas

wastewater samples (Sample 01 Liquid, Sample 02 Liquid, and Sample 03 Liquid) for

concentrations of Cl, Br, S, O, Li, B, Na, K, Mg, Ca, Sr, Ba, Al, Fe, Mn, S, Cr, Ni, Cu, Zn, As, Cd,

and Pb elements. The laboratories that do not have the appropriate equipment or technique to

perform a quantitative analysis on all the elements were allowed to only report within their

capabilities. Also, each laboratory was asked to submit their results to an anonymous online portal

with an attachment of the sample preparation details (dilution factor, precipitation or evaporation

details, etc.), the list of utilized equipment and methodology, uncertainty levels, and calibration

descriptions [1].

The laboratories involved in this study utilized a variety of methods, such as inductively

coupled plasma with optical emission spectrometry (ICP-OES), inductively coupled plasma with

mass spectrometry (ICP-MS), triple quadrupole inductively coupled plasma with mass

spectrometry (ICP-MS/MS), direct current plasma (DCP), X-ray fluorescence (XRF), ion

64

chromatography (IC), and neutron activation analysis (NAA). Some of the commercial laboratories

were previously experienced in analyzing oil and gas wastewater and issued a certificate of analysis

for regulatory applications. Thus, the main objectives of this study were 1) to collect and compare

the trace element concentration values of high salinity oil and gas wastewater, 2) to evaluate the

quality of results of each technique, 2) identify the most suitable application for oil and gas

wastewater characterizations, 4) evaluate the analytical accuracy of detection of target elements

[1].

First, the submitted data were pared down by excluding all zero and below detection limit

values. Then, the numerical values were processed and evaluated using the nonparametric statistical

method that is common for inter-laboratory comparison studies within the United States Geological

Survey (USGS). This method is more resistant to outlier results since it is based not on mean

(average) but, on Quartile 1 (Q1), median, and Quartile 3 (Q3) values. After determining all Q1,

median, and Q3 values for each analyte concertation in each sample, the uncertainty magnitudes

(standard deviation (F-pseudosigma)) were calculated using Equation 7.3.1.

F − pseudosigma =Q3−Q1

1.349 (7.3.1.)

Where 1.349 is a constant that represents the standard deviations necessary to include the

interquartile range data (i.e., Q3-Q1). The median value was claimed as the most probable value

(MPV), in cases where seven or more values were reported per sample and the F-pseudosigma

magnitude was not higher than the median itself. For analytes with only six or five reported values,

the median and F-pseudosigma quantities were evaluated along with the Q1 and Q3 concentration

[1]. Statistical determination was not performed for analytes with less than five reported values and

are marked as not calculable (n.c.) in Table 7.3.1. The most probable value (MPV), F-pseudosigma

(F), lower and upper quartile magnitudes for each analyte are also demonstrated in Table 7.3.1.

65

Table 7.3.1. A summary of inter-laboratory study. All values are represented in mg/l [1].

Analyte Sample 01 Liquid Sample 02 Liquid Sample 03 Liquid

MPV Q1-Q3 F MPV Q1-Q3 F MPV Q1-Q3 F

Br 746 652 - 773 90.4 1270 1180 - 1440 189 1890 1630 - 2060 320

Cl 65600 63900 - 68300 3300 117000 113000 - 120000 5470 176000 160000 - 180000 15000

SO4 n.c. n.c. n.c. n.c. n.c. n.c. 170* 130 – 172* 33.0*

Na 27000 24900 - 28600 2710 47500 43600 - 49300 4260 66850 64600 - 68900 3170

K 336 276 - 383 79.3 716 621 - 765 107 2190 1770 - 2310 402

Mg 1230 1200 - 1300 69.3 2168 2100 - 2270 127 3100 2990 - 3130 104

Ca 10000 9280 - 10200 686 19800 18600 - 20600 1480 31400 30000 - 33200 2350

Sr 2160 2130 - 2200 49.7 3710 3580 - 3940 270 1540 1410 - 1620 156

Ba 659 641 - 690 37.2 1320 1280 - 1380 72.8 6.12 6.07 - 6.33 0.195

Li 32.1 30.3 - 34.3 3 50.3 48 - 51 2.19 71.7 68 - 74.2 4.6

B 5 3.95 - 5.09 0.85 7 6.76 - 8.05 0.95 15.3 14.7 - 16 0.999

Al n.c. n.c. n.c. n.c. n.c. n.c. n.c. n.c. n.c.

Fe 64.8 58.7 - 69 7.61 94.9 85.8 - 98.5 9.44 169 158 - 181 17

Mn 6.1 5.75 - 6.7 0.7 14.4 13.7 - 14.9 0.93 47.8 41.5 - 48.3 5.06

In order to evaluate the performance of the NAA method in the characterization of oil and

gas wastewater samples, the most probable values (MPVs) that obtained from the inter-laboratory

study were compared with NAA measured values (MVs). However, as discussed in Section 7.1,

the NAA method is not applicable to the quantitative analysis of all analytes listed in Table 7.3.1,

therefore the comparison is conducted only among NAA suitable elements. To make the

comparison more convenient, the trace element concentration and standard deviation values were

converted from weight percent (wt%) to milligram per liter (mg/l) and arranged in Table 7.3.2.

Table 7.3.2. The concentration and standard deviation values of some trace elements

measured using the NAA method.

Analyte Sample 01 Liquid Sample 02 Liquid Sample 03 Liquid

MV (mg/l) Error MV (mg/l) Error MV (mg/l) Error

Sodium 24760.8 740.8 39596.5 2059.6 31271.4 1806.3

Potassium BDL --- BDL --- 2203.0 114.7

Calcium 7434.7 348.9 13471.8 670.1 28543.3 1931.5

Strontium 2162.9 51.4 4083.8 95.2 669.3 52.6

Barium 622.4 31.1 1221.8 66.0 7.2 0.9

Iron 54.3 1.4 82.8 5.9 165.1 13.2

Manganese 5.4 0.9 13.1 1.6 51.5 6.5

66

Figure 7.3.1 graphically shows the comparison of manganese concentrations in all three

wastewater samples, which are determined by NAA method and inter-laboratory study (MPV).

The graphical comparison of concentrations of the remaining elements can be found in

Appendix D.

Figure 7.3.1. A comparison of manganese concentrations in oil and gas wastewater samples

in a graphical manner.

In addition, the percent difference between MPV and NAA measured values were also

calculated numerically using the following equation.

Difference (%) = [NAA MV−MPV

MPV] × 100% (7.3.2)

The results of calculation are shown in Table 7.3.3.

6.1 5.4

14.4 13.1

47.851.5

0.0E+00

1.0E+01

2.0E+01

3.0E+01

4.0E+01

5.0E+01

6.0E+01

Con

centr

atio

n (

mg/l

)

Sample 01 Liquid Sample 02 Liquid Sample 03 Liquid

Manganese

Most probable value (MPV) NAA mesured value

67

Table 7.3.3. The percent difference magnitudes between MPV and NAA measured

values.

Element Sample 01 Liquid Sample 02 Liquid Sample 03 Liquid

Sodium 8.3 16.6 53.2

Potassium --- --- 0.6

Calcium 25.7 32.0 9.1

Strontium 0.1 10.1 56.5

Barium 5.6 7.4 16.9

Iron 16.1 12.8 2.3

Manganese 11.6 8.7 7.8

If the quality evaluation ranges are set as <20% is acceptable, 20%-40% questionable, and

>40% unacceptable, the NAA provided two questionable (calcium) and two unacceptable

(strontium and sodium) out of 19 values. The barium, iron, and manganese concentration values

agreed with most probable values, since all these elements have the best combination of NAA-

relevant half-time period, relatively high gamma ray energy and intensity.

68

Chapter 8

Conclusion and Future Works

This research determined the multi-elemental characterization of fifteen solid (soil and

sediment) and seven liquid (wastewater) hydraulic fracturing samples collected from wellbores

within Pennsylvania. The characterization was performed using the comparative neutron activation

analysis (CNAA) method for several reasons, such as availability of standard reference materials

(SRM) for comparison, less sensitivity to changes in measurement parameters, similar structure

and geometries of the SRM, powder (soil and sediment), and small crystals (dried wastewater)

samples, as well as less complexity in data analysis. The Buffalo River Sediment and Montana Soil

SRMs obtained from National Institute of Standards and Technology (NIST) were selected as the

standard comparators, since most of the test samples are soil and sediment. In addition, these SRMs

contain almost all of 23 target elements (except chlorine in Montana Soil) for this study, which

includes Al, B, Na, K, Mg, Ca, Sr, Ba, Cr, Mn, Fe, Ni, Cu, Zn, As, Np (uranium activation product),

Pb, Pa (thorium activation product), Cl, Br, Se, Ag, and Hg.

In order to ensure safety and high efficiency with respect to the neutron exposure rate and

time, the neutron flux magnitudes were measures at different axial levels along Dry Tube 1 (DT1)

for the PSBR Core 58 loading. The thermal and resonance (epithermal) neutron flux profiles within

DT1 were determined by irradiating and counting of the bare and cadmium covered aluminum-

gold (0.112% gold) wires. The peak of the thermal and resonance fluxes was determined at 10.25

inches above the DT1 bottom and reached magnitudes of 1.09x1013 n/cm2s and 7.6x1011 n/cm2s

respectively.

Further, using the determined neutron flux values, Dr. Dağistan Şahin’s Activity Predictor

program [28] was used to develop an experimental design in terms of sample preparation,

69

irradiation and decay times, and radiation counting durations. All solid fracking samples and SRMs

were weighed, placed into ampoules made from the Heraeus Suprasil 310 quartz tubes, and then

sealed. The liquid samples were dehydrated under the heat lamps until they turned into crystals,

which were subsequently also weighed and encapsulated in the same manner. Such encapsulation

allows to preserve the integrity of the sample matrices after prolonged irradiation and to avoid

internal contamination of the irradiation fixture. Moreover, pure quartz will not be easily activated,

so it negligibly contributes to the total activity of the samples and to the background radiation

during the counting. Then, all samples were placed into two small aluminum buckets following the

pattern shown in Figure 6.2.4. There were two short and two long irradiations conducted for this

research. The short irradiations designed for quantitative analysis of the elements with short-lived

(T1/2 ≤ 15 hours) radionuclides. The short irradiations lasted for 6 minutes at the reactor power of

110 kW and 10 kW with Bucket #1 and Bucket #2 respectively. Long irradiations aimed to analyze

the intermedium-lived (15 hours < T1/2 < 7 days), and long-lived (T1/2 ≥ 7 days) radioisotopes.

Each long irradiation was executed at the reactor power 800 kW for 10 hours. The long irradiations

were split for two reactor operation days, and this fact would not make any difference, since test

samples and comparators would undergo the same neutron fluence and be analyzed using the

CNAA method. To collect data on the short-lived radionuclides, the samples were counted for 15

minutes using the HPGe detector located in a lead cave. Due to high deadtime caused by high

sodium concentrations, the samples were counted in multiple rounds with a range of the decay

period from eight hours to two days. The intermedium-lived radioisotopes were counted for 50

minutes at decay period range of 6-12 days. To acquire a sufficient data on the long-lived

radionuclides, each sample was measured for 3 hours after 22-28 days elapsed from the irradiation.

After collecting gamma ray spectra from all measurements, they were analyzed. The data

analysis started from selecting the NAA suitable isotopes for each target element based on the

natural abundance, neutron absorption cross-section, half-life period, and gamma ray emission

70

intensity and energy characteristics. After this step, the initial list of elements of interest was

reduced to 14 elements, since nine of the desired elements did not have any NAA-relevant isotopes.

During further analysis it was noticed that the photo-peaks with energy corresponding to Eu-152m,

La-140, Rb-86, Sc-46, and Cs-134 isotopes often present in gamma ray spectra of the samples, so

the list of the target elements was updated again. Finally, all samples were analyzed on the

concentration values of 19 trace elements given in Table 7.1.1. After determining the

concentrations of trace elements in each sample, the error analysis was conducted via mass,

measured activity, and decay time uncertainties. The experimentally measured trace element

concentration and error values in weigh percent (wt%) are shown in Tables 7.1.3 and 7.1.4.

Moreover, the elemental concentrations for each individual sample can be found in Table 7.1.2 and

in Appendix D.

To check and verify experimentally measured values, a quality control analysis was

performed using both SRMs. As a result of this analysis, the magnitude of the error in chlorine and

cesium concentrations was not verified, therefore their values were declared non-certified. Arsenic,

mercury, strontium, and selenium demonstrated a high (>20%) percentage of error in certain

measurements (Tables 7.2.1 and 7.2.2). The high error in arsenic concentration is probably

associated with high deadtime. After a short irradiation, the Bucket #2 contents were measured

only once because of the limited time interval caused by the scheduled renovation of the PSBR. In

addition, a high error in selenium concentration can be explained by the lack of irradiation time.

Mercury and strontium may have needed one more counting round with a decay period over 30

days.

Another accuracy check was carried out by comparison of the NAA measured values with

the most probable values (MPV) obtained through inter-laboratory study. In this study participated

15 different laboratories from the US, Canada, and Germany, which used a variety of methods to

determine the trace element concentration of three oil and gas wastewaters. The collected results

71

were evaluated and analyzed using the nonparametric statistical method that allows to identify the

most probable concentration value and approximate standard deviation (Table 7.3.1). According to

the interlaboratory study, the NAA method seems to be provide less accurate results than the

inductively coupled plasma (ICP) method, which traditionally is used for determining trace element

concentrations in oil and gas wastewaters [1]. The comparison of NAA measured values at the

RSEC and the most probable values obtained through inter-laboratory study showed the percent

difference between 0.1% and 56.6%. The relatively large differences in these values might be

caused by several reasons. First, all three samples are liquid wastewaters, which contain a high

concentration of hydride even after the dehydration process. Secondly, due to the evaporation and

absorption of water from the air, it was challenging to weigh both liquid and dried crystalline

samples; thus, the mass error might be high. Thirdly, to avoid a large amount of water absorption,

the crystals were not pulverized, which could cause heterogeneity of the test samples. Finally,

during dehydration, liquid samples were placed from vial to vial, which may lead to a loss of

accuracy.

As the results of this research showed, the NAA and in particular CNAA is a satisfactory

method for conducting quantitative analysis of particular (NAA suitable) elements in the hydraulic

fracturing soil, sediment, and wastewater samples. This technique can be useful as a secondary

analysis to double check results of the prime characterization conducted using the traditional

methods. Moreover, the interlaboratory comparison was conducted for only three liquid samples

and was not the case for any solid samples. Thus, there is a possibility that the NAA may perform

better than other conventional methods in determining the trace element concentrations in fracking

soil and sediments.

Based on the results obtained in this study, the experiment design can be improved for

future work in terms of irradiation, decay, and radiating counting periods. The optimization of the

experiment will reduce detection limits, and subsequently provide more numerical values. Also,

72

the preparation of the liquid samples can be advanced by eliminating mistakes that were made

during the dehydration process in this work.

The range of target elements can be expanded by adding more elements that are typically

challenging for NAA determination. The pneumatic transport system at the Radiation Science and

Engineering Center (RSEC) can be involved to irradiate fracking samples for a short period at low

reactor power, which allows counting samples with short decay period and analyzed very short-

lived radionuclides, such as Mg-27, Br-80, Ag-110, etc. Due to deposition of the partial energy

caused by Compton scattering of gamma rays in the detector, the Compton continuum appears in

the gamma ray spectrum, which obscures most of the low energy photopeaks; thus, it is challenging

to analyze the radioisotopes with lower energy gamma ray emission. However, the Radionuclear

Applications Laboratory (RAL) at the RSEC is equipped with a Compton Suppression System that

reduces the intensity of Compton continuum and more accurately identify the low-energy

photopeaks located within the Compton plateau [23]. This analysis may allow to characterize

radioisotopes such as Mg-28, Np-239, etc.

73

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76

Appendix A

Physical Specifications

The HPGe detector dimensional characteristics provided by the manufacturer are shown

in Figure A-1.

Figure A-1. The HPGe detector dimensions provided by the manufacturer

77

Appendix B

NIST Certificates

The following is the NIST certificate for the Buffalo River Sediment that was used as a

reference standard material for CNAA.

78

79

80

The following is the NIST certificate for the Montana Soil that was used for the quality

control of the results.

81

82

83

84

Appendix C

Activity Prediction Results

The calculated activity and dose rate values at the end of irradiation are demonstrated in

Tables C-1 through C-4. The values were obtained for Buffalo River Sediment standard reference

material using the Activity Predictor program developed by Dr. Dağistan Şahin. Due to the length

of used tweezers the gamma exposure rate was calculated for 1, 10, and 30 cm distances. The mass

of the sample was given as 0.075 g.

Table C-1. Calculated activities and dose rates for the end of short irradiation of Bucket

#1 content after a decay period of 48 hours.

Isotope Mass (g) Activity (mCi) Gamma Exposure Rate (mrem/hr)

Decay 48 h 1 cm 10 cm 30 cm

Na-24 4.58E-03 5.70E-02 1.61E+04 1.61E+02 1.79E+01

Ca-47 7.80E-08 7.03E-06 3.23E-02 3.23E-04 3.58E-05

Mn-54 1.79E-04 2.04E-06 9.53E-03 9.53E-05 1.06E-05

Cr-51 1.79E-04 1.59E-04 9.79E-02 9.79E-04 1.09E-04

Mn-56 2.83E-03 8.27E-06 6.08E-02 6.08E-04 6.76E-05

Fe-59 8.63E-06 1.58E-04 8.57E-01 8.57E-03 9.52E-04

P-32 7.49E-05 9.92E-04 2.64E+02 2.64E+00 2.93E-01

K-42 1.01E-04 1.76E-02 2.09E+01 2.09E-01 2.32E-02

Si-31 6.76E-04 2.58E-06 4.84E-01 4.84E-03 5.37E-04

Sc-46 2.74E-05 2.22E-04 2.10E+00 2.10E-02 2.33E-03

Sc-47 2.50E-05 5.00E-06 2.33E-03 2.33E-05 2.59E-06

Sc-48 2.53E-04 1.04E-06 1.60E-02 1.60E-04 1.78E-05

Sb-122 1.63E-07 6.75E-05 1.44E-01 1.44E-03 1.60E-04

Sb-124 1.21E-07 2.37E-06 1.92E-02 1.92E-04 2.14E-05

As-76 1.76E-06 8.36E-04 1.53E+00 1.53E-02 1.70E-03

Ba-131 3.29E-08 5.87E-06 1.63E-02 1.63E-04 1.82E-05

Ba-133m 3.14E-08 1.76E-06 6.94E-04 6.94E-06 7.71E-07

Ba-135m 7.50E-07 8.21E-06 4.78E-03 4.78E-05 5.31E-06

Cd-115 7.43E-08 1.32E-06 2.97E-03 2.97E-05 3.30E-06

Co-60 1.05E-06 1.26E-05 1.43E-01 1.43E-03 1.59E-04

Co-58 1.05E-06 1.46E-07 7.69E-04 7.69E-06 8.54E-07

Cu-64 5.12E-06 1.70E-03 2.06E+00 2.06E-02 2.29E-03

Hg-203 3.29E-08 6.01E-07 0.00E+00 0.00E+00 0.00E+00

Zn-65 1.60E-05 2.60E-05 1.23E-01 1.23E-03 1.37E-04

Zn-69m 6.18E-06 3.68E-05 8.08E-02 8.08E-04 8.98E-05

S-35 1.26E-04 3.82E-04 0.00E+00 0.00E+00 0.00E+00

Br-80m 2.66E-07 7.08E-07 4.35E-03 4.35E-05 4.84E-06

Br-82 2.59E-07 7.95E-05 9.92E-01 9.92E-03 1.10E-03

Ce-137 1.03E-08 1.09E-06 6.73E-04 6.73E-06 7.48E-07

Ce-137m 1.03E-08 6.65E-07 3.23E-04 3.23E-06 3.59E-07

85

Ce-141 4.78E-06 2.08E-05 8.10E-03 8.10E-05 9.00E-06

Ce-143 5.98E-07 3.96E-05 6.38E-02 6.38E-04 7.09E-05

Cs-134 4.50E-07 4.96E-06 3.77E-02 3.77E-04 4.18E-05

Dy-165 1.27E-07 4.87E-07 2.08E-05 2.08E-07 2.31E-08

Eu-152 4.66E-08 1.34E-05 6.99E-02 6.99E-04 7.77E-05

Eu-152m 4.66E-08 2.63E-03 3.16E+00 3.16E-02 3.51E-03

Eu-154 5.09E-08 1.18E-06 6.47E-03 6.47E-05 7.19E-06

Ga-72 4.49E-07 5.56E-06 2.27E+00 2.27E-02 2.53E-03

La-140 2.17E-06 1.37E-03 1.37E+01 1.37E-01 1.52E-02

Lu-176m 4.38E-08 1.15E-07 7.02E-06 7.02E-08 7.80E-09

Lu-177 1.17E-09 6.11E-05 9.09E-03 9.09E-05 1.01E-05

Rb-86 5.41E-06 5.59E-05 2.52E-02 2.52E-04 2.80E-05

Se-75 7.43E-10 1.43E-07 2.52E-04 2.52E-06 2.80E-07

Sr-85 5.46E-08 2.33E-07 1.32E-03 1.32E-05 1.47E-06

Sr-89 8.05E-06 3.68E-07 1.60E-07 1.60E-09 1.78E-10

Sm-153 1.34E-07 1.78E-03 7.30E-01 7.30E-03 8.11E-04

Yb-169 2.73E-10 5.88E-06 9.35E-03 9.35E-05 1.04E-05

Yb-175 6.68E-08 3.04E-04 5.66E-02 5.66E-04 6.29E-05

Zr-95 3.91E-06 1.25E-06 4.46E-03 4.46E-05 4.96E-06

Zr-97 6.30E-07 1.05E-06 4.19E-03 4.19E-05 4.65E-06

Table C-1 values were calculated for 6-minute irradiation at the reactor power of 110

kW.

86


#2 content after a decay period of 48 hours.



Na-24 4.58E-03 5.18E-03 1.46E+03 1.46E+01 1.62E+00

Ca-47 7.80E-08 6.39E-07 2.93E-03 2.93E-05 3.26E-06

Mn-54 1.79E-04 1.86E-07 8.66E-04 8.66E-06 9.63E-07

Cr-51 1.79E-04 1.44E-05 8.90E-03 8.90E-05 9.89E-06

Mn-56 2.83E-03 7.52E-07 5.53E-03 5.53E-05 6.14E-06

Fe-59 8.63E-06 1.44E-05 7.79E-02 7.79E-04 8.65E-05

P-32 7.49E-05 9.02E-05 2.40E+01 2.40E-01 2.66E-02

K-42 1.01E-04 1.60E-03 1.90E+00 1.90E-02 2.11E-03

Si-31 6.76E-04 2.34E-07 4.40E-02 4.40E-04 4.88E-05

Sc-46 2.74E-05 2.02E-05 1.91E-01 1.91E-03 2.12E-04

Sc-47 2.50E-05 4.55E-07 2.12E-04 2.12E-06 2.36E-07

Sb-122 1.63E-07 6.14E-06 1.31E-02 1.31E-04 1.46E-05

Sb-124 1.21E-07 2.16E-07 1.75E-03 1.75E-05 1.94E-06

As-76 1.76E-06 7.60E-05 1.39E-01 1.39E-03 1.55E-04

Ba-131 3.29E-08 5.34E-07 1.49E-03 1.49E-05 1.65E-06

Ba-133m 3.14E-08 1.60E-07 6.31E-05 6.31E-07 7.01E-08

Ba-135m 7.50E-07 7.47E-07 4.34E-04 4.34E-06 4.83E-07

Cd-115 7.43E-08 1.20E-07 2.70E-04 2.70E-06 3.00E-07

Co-60 1.05E-06 1.15E-06 1.30E-02 1.30E-04 1.44E-05

Cu-64 5.12E-06 1.54E-04 1.87E-01 1.87E-03 2.08E-04

Zn-65 1.60E-05 2.36E-06 1.12E-02 1.12E-04 1.24E-05

Zn-69m 6.18E-06 3.34E-06 7.34E-03 7.34E-05 8.16E-06

S-35 1.26E-04 3.47E-05 0.00E+00 0.00E+00 0.00E+00

Br-82 2.59E-07 7.23E-06 9.02E-02 9.02E-04 1.00E-04

Ce-141 4.78E-06 1.89E-06 7.36E-04 7.36E-06 8.18E-07

Ce-143 5.98E-07 3.60E-06 5.80E-03 5.80E-05 6.44E-06

Cs-134 4.50E-07 4.51E-07 3.42E-03 3.42E-05 3.80E-06

Eu-152 4.66E-08 1.21E-06 6.36E-03 6.36E-05 7.06E-06

Eu-152m 4.66E-08 2.39E-04 2.88E-01 2.88E-03 3.19E-04

Eu-154 5.09E-08 1.07E-07 5.88E-04 5.88E-06 6.54E-07

Ga-72 4.49E-07 5.05E-07 2.07E-01 2.07E-03 2.30E-04

La-140 2.17E-06 1.24E-04 1.25E+00 1.25E-02 1.39E-03

Lu-177 1.17E-09 5.56E-06 8.26E-04 8.26E-06 9.18E-07

Rb-86 5.41E-06 5.08E-06 2.29E-03 2.29E-05 2.55E-06

Sm-153 1.34E-07 1.62E-04 6.64E-02 6.64E-04 7.38E-05

Yb-169 2.73E-10 5.35E-07 8.50E-04 8.50E-06 9.45E-07

Yb-175 6.68E-08 2.77E-05 5.14E-03 5.14E-05 5.71E-06

Zr-95 3.91E-06 1.13E-07 4.06E-04 4.06E-06 4.51E-07

Table C-2 values were calculated for irradiation at the reactor power of 10 kW for 6

minutes.

87


#1 and Bucket #2 content after a decay period of 192 hours.



Ca-47 7.80E-08 2.81E-06 1.528E-02 1.528E-04 1.697E-05

Mn-54 1.79E-04 2.01E-06 9.324E-03 9.324E-05 1.036E-05

Cr-51 1.79E-04 1.37E-04 2.433E-02 2.433E-04 2.704E-05

Fe-59 8.63E-06 1.44E-04 8.922E-01 8.922E-03 9.913E-04

P-32 7.49E-05 7.39E-04 1.968E+02 1.968E+00 2.187E-01

Sc-46 2.74E-05 2.12E-04 2.285E+00 2.285E-02 2.538E-03

Sc-47 2.50E-05 1.45E-06 7.722E-04 7.722E-06 8.580E-07

Sc-48 2.53E-04 1.05E-07 1.867E-03 1.867E-05 2.074E-06

Sb-122 1.63E-07 1.48E-05 3.794E-02 3.794E-04 4.215E-05

Sb-124 1.21E-07 2.22E-06 2.121E-02 2.121E-04 2.357E-05

Ba-131 3.29E-08 4.08E-06 1.341E-02 1.341E-04 1.490E-05

Ba-133 3.14E-08 1.32E-05 4.007E-02 4.007E-04 4.452E-05

Cd-109 2.30E-09 1.25E-06 2.372E-03 2.372E-05 2.635E-06

Cd-115 7.43E-08 2.04E-07 2.369E-04 2.369E-06 2.632E-07

Cd-115m 7.43E-08 1.24E-05 2.163E-03 2.163E-05 2.404E-06

Co-60 1.05E-06 1.19E-05 1.529E-01 1.529E-03 1.699E-04

Co-58 1.05E-06 1.38E-07 7.496E-04 7.496E-06 8.328E-07

Hg-203 3.29E-08 5.50E-07 7.144E-04 7.144E-06 7.938E-07

Tl-204 2.66E-08 2.79E-05 1.645E-04 1.645E-06 1.828E-07

Zn-65 1.60E-05 5.36E-03 1.645E+01 1.645E-01 1.827E-02

S-35 1.26E-04 5.53E-06 0.000E+00 0.000E+00 0.000E+00

Ce-139 1.35E-08 1.93E-05 2.452E-02 2.452E-04 2.724E-05

Ce-141 4.78E-06 1.83E-05 8.289E-03 8.289E-05 9.210E-06

Cs-134 5.98E-07 4.93E-06 4.322E-02 4.322E-04 4.802E-05

Dy-159 4.50E-07 2.07E-05 9.166E-03 9.166E-05 1.018E-05

Eu-152 4.66E-08 1.34E-05 8.620E-02 8.620E-04 9.578E-05

Eu-154 5.09E-08 1.18E-06 7.880E-03 7.880E-05 8.756E-06

La-140 2.17E-06 9.71E-05 1.136E+00 1.136E-02 1.262E-03

Lu-177 1.17E-09 2.77E-05 5.015E-03 5.015E-05 5.572E-06

Lu-177m 1.17E-09 7.33E-06 4.011E-02 4.011E-04 4.456E-05

Rb-86 5.41E-06 6.48E-05 3.206E-02 3.206E-04 3.562E-05

Se-75 7.43E-10 1.73E-07 3.508E-04 3.508E-06 3.898E-07

Sr-85 5.46E-08 3.04E-07 8.691E-04 8.691E-06 9.657E-07

Sr-89 8.05E-06 4.30E-07 1.876E-07 1.876E-09 2.084E-10

Sm-145 1.56E-08 5.48E-06 4.587E-03 4.587E-05 5.096E-06

Sm-153 1.34E-07 2.84E-04 1.368E-01 1.368E-03 1.520E-04

Sn-113 6.91E-09 6.82E-06 8.256E-03 8.256E-05 9.174E-06

Sn-117m 1.04E-07 6.68E-06 1.128E-02 1.128E-04 1.254E-05

Sn-119m 1.73E-07 2.03E-06 1.824E-03 1.824E-05 2.027E-06

Sn-123 3.30E-08 9.68E-06 3.522E-04 3.522E-06 3.913E-07

Sn-125 4.13E-08 2.20E-06 3.872E-03 3.872E-05 4.302E-06

Yb-169 2.73E-10 6.24E-06 1.211E-02 1.211E-04 1.345E-05

Yb-175 6.68E-08 9.47E-05 2.082E-02 2.082E-04 2.314E-05

Zr-95 3.91E-06 1.22E-06 5.025E-03 5.025E-05 5.583E-06

88

Table C-3 and Table C-4 values are predicted for 10-hour irradiation at the rector power

of 800 kW.


#1 and Bucket #2 content after a decay period of 552 hours.



Ca-47 7.80E-08 2.85E-07 1.547E-03 1.547E-05 1.719E-06

Mn-54 1.79E-04 1.95E-06 9.019E-03 9.019E-05 1.002E-05

Cr-51 1.79E-04 9.39E-05 1.672E-02 1.672E-04 1.858E-05

Fe-59 8.63E-06 0.000114 7.063E-01 7.063E-03 7.848E-04

P-32 7.49E-05 0.000357 9.499E+01 9.499E-01 1.055E-01

Sc-46 2.74E-05 0.000187 2.018E+00 2.018E-02 2.242E-03

Sb-124 1.21E-07 1.86E-06 1.785E-02 1.785E-04 1.983E-05

Ba-131 3.29E-08 1.65E-06 5.430E-03 5.430E-05 6.033E-06

Ba-133 3.14E-08 1.31E-05 3.996E-02 3.996E-04 4.440E-05

Cd-109 2.30E-09 1.23E-06 2.319E-03 2.319E-05 2.577E-06

Cd-115m 7.43E-08 9.79E-06 1.713E-03 1.713E-05 1.904E-06

Co-60 1.05E-06 1.04E-05 1.343E-01 1.343E-03 1.492E-04

Co-58 1.05E-06 1.19E-07 6.473E-04 6.473E-06 7.192E-07

Hg-203 3.29E-08 4.4E-07 5.716E-04 5.716E-06 6.351E-07

Tl-204 2.66E-08 2.77E-05 1.633E-04 1.633E-06 1.814E-07

Zn-65 1.60E-05 0.005133 1.576E+01 1.576E-01 1.751E-02

S-35 1.26E-04 4.91E-06 0.000E+00 0.000E+00 0.000E+00

Ce-139 1.35E-08 1.79E-05 2.274E-02 2.274E-04 2.526E-05

Ce-141 4.78E-06 1.33E-05 6.021E-03 6.021E-05 6.690E-06

Cs-134 5.98E-07 4.87E-06 4.263E-02 4.263E-04 4.736E-05

Dy-159 4.50E-07 1.93E-05 8.529E-03 8.529E-05 9.477E-06

Eu-152 4.66E-08 1.34E-05 8.602E-02 8.602E-04 9.558E-05

Eu-154 5.09E-08 1.17E-06 7.854E-03 7.854E-05 8.727E-06

Lu-177 1.17E-09 5.8E-06 1.050E-03 1.050E-05 1.166E-06

Lu-177m 1.17E-09 6.87E-06 3.759E-02 3.759E-04 4.177E-05

Rb-86 5.41E-06 3.71E-05 1.836E-02 1.836E-04 2.040E-05

Se-75 7.43E-10 1.58E-07 3.216E-04 3.216E-06 3.574E-07

Sr-85 5.46E-08 2.59E-07 7.404E-04 7.404E-06 8.226E-07

Sr-89 8.05E-06 3.5E-07 1.527E-07 1.527E-09 1.697E-10

Sm-145 1.56E-08 5.31E-06 4.449E-03 4.449E-05 4.943E-06

Sn-113 6.91E-09 6.23E-06 7.543E-03 7.543E-05 8.381E-06

Sn-117m 1.04E-07 3.18E-06 5.370E-03 5.370E-05 5.966E-06

Sn-119m 1.73E-07 1.96E-06 1.760E-03 1.760E-05 1.956E-06

Sn-123 3.30E-08 8.93E-06 3.250E-04 3.250E-06 3.611E-07

Sn-125 4.13E-08 7.48E-07 1.317E-03 1.317E-05 1.463E-06

Yb-169 2.73E-10 4.51E-06 8.750E-03 8.750E-05 9.722E-06

Yb-175 6.68E-08 7.9E-06 1.737E-03 1.737E-05 1.930E-06

Zr-95 3.91E-06 1.04E-06 4.272E-03 4.272E-05 4.747E-06

89

Appendix D

Analysis Results

The trace element concentrations for each of the tested hydraulic fracturing samples are

shown in Tables D-1 through D-21. The <*> sign in the first column represents non-certified

values that have not passed quality control; also, the BDL stands for below detection limit.

Table D-1. Experimentally determined trace element concentrations of HR SaH sample

using NAA method.

Element Nuclide Half-life

(h)

Short

(wt%) Error

Medium

(wt%) Error

Long

(wt%) Error

Chlorine* Cl-39 0.9367 BDL --- Manganese Mn-56 2.5789 BDL --- Europium Eu-152m 9.3116 BDL --- Potassium K-42 12.355 8.94E-02 1.01E-01 Sodium Na-24 14.997 3.08E+01 3.27E-02 Arsenic As-76 26.24 4.93E-02 5.32E-03

Lanthanum La-140 40.2852 BDL --- BDL --- BDL ---

Calcium Ca-47 108.864 1.40E+00 2.31E-01 1.40E+00 1.01E-01

Barium Ba-131 276 7.53E-03 1.60E-03 7.43E-03 4.44E-04

Rubidium Rb-86 447.408 BDL --- 1.01E-04 2.27E-05

Protactinium Pa-233 647.4 BDL --- BDL ---

Chromium Cr-51 664.896 BDL --- BDL ---

Iron Fe-59 1067.88 BDL --- 6.90E-04 4.23E-04

Mercury Hg-203 1118.26 BDL --- BDL ---


Scandium Sc-46 2010.96 3.82E-07 2.50E-07 1.32E-07 4.17E-08

Selenium Se-75 2874.72 BDL --- BDL ---

Zinc Zn-65 5854.32 BDL --- 6.76E-05 1.33E-05

Cesium Cs-134 18091.1 BDL --- 2.28E-05 6.13E-07

90


NAA method.


(h)

Short

(wt%) Error

Medium

(wt%) Error

Long

(wt%) Error

Chlorine* Cl-39 0.9367 2.70E-02 6.41E-03 Manganese Mn-56 2.5789 1.99E-02 3.62E-02 Europium Eu-152m 9.3116 BDL --- Potassium K-42 12.355 1.48E-01 4.01E-02 Sodium Na-24 14.997 1.24E+00 4.45E-03 Arsenic As-76 26.24 BDL ---


Calcium Ca-47 108.864 1.44E+00 8.15E-02 1.13E+00 9.41E-02

Barium Ba-131 276 4.45E+01 2.97E-02 4.11E+01 2.43E-02

Rubidium Rb-86 447.408 2.72E-04 3.82E-05 5.64E-04 2.86E-05


Chromium Cr-51 664.896 1.65E-02 1.03E-03 1.26E-02 3.72E-04

Iron Fe-59 1067.88 1.28E-02 3.96E-03 1.43E-02 9.86E-04



Scandium Sc-46 2010.96 5.36E-07 1.78E-07 6.33E-07 2.62E-10


Zinc Zn-65 5854.32 BDL --- 4.67E-05 2.77E-05

Cesium Cs-134 18091.1 3.20E-05 3.71E-06 7.69E-06 1.16E-06


NAA method.


(h)

Short

(wt%) Error

Medium

(wt%) Error

Long

(wt%) Error

Chlorine* Cl-39 0.9367 3.17E-02 1.67E-02 Manganese Mn-56 2.5789 5.68E-03 7.68E-03 Europium Eu-152m 9.3116 BDL --- Potassium K-42 12.355 2.50E-01 5.92E-02 Sodium Na-24 14.997 1.01E+00 4.04E-03 Arsenic As-76 26.24 BDL ---


Calcium Ca-47 108.864 1.19E+00 8.86E-02 1.30E+00 9.69E-02

Barium Ba-131 276 4.53E+01 3.04E-02 4.23E+01 2.51E-02

Rubidium Rb-86 447.408 3.75E-04 4.69E-05 4.30E-04 1.84E-05

Protactinium Pa-233 647.4 BDL --- 7.59E-06 2.35E-08


Iron Fe-59 1067.88 1.17E-02 1.60E-03 1.14E-02 8.79E-04


Strontium Sr-85 1556.38 6.56E+00 1.20E-02 8.10E-00 8.52E-03

Scandium Sc-46 2010.96 4.70E-07 1.29E-07 4.87E-07 5.68E-08


Zinc Zn-65 5854.32 1.66E-03 9.56E-05 4.04E-04 2.64E-05

Cesium Cs-134 18091.1 BDL --- BDL ---

91


sample using NAA method.


(h)

Short

(wt%) Error

Medium

(wt%) Error

Long

(wt%) Error

Chlorine* Cl-39 0.9367 4.88E-02 1.03E-02 Manganese Mn-56 2.5789 BDL --- Europium Eu-152m 9.3116 1.53E-04 4.06E-06 Potassium K-42 12.355 1.42E+00 3.63E-02 Sodium Na-24 14.997 5.21E-02 9.37E-04 Arsenic As-76 26.24 2.64E-03 9.58E-05



Barium Ba-131 276 5.63E-02 2.61E-03 4.85E-02 1.79E-03

Rubidium Rb-86 447.408 7.45E-03 1.55E-04 6.99E-03 9.65E-05


Chromium Cr-51 664.896 6.02E-04 1.34E-05 1.36E-02 2.14E-04

Iron Fe-59 1067.88 8.09E-01 7.33E-03 1.16E+00 6.02E-03

Mercury Hg-203 1118.26 BDL --- 2.63E-04 2.37E-05

Strontium Sr-85 1556.38 BDL --- BDL ---

Scandium Sc-46 2010.96 1.64E-03 3.81E-06 1.59E-03 2.16E-06

Selenium Se-75 2874.72 2.89E-04 2.24E-05 5.58E-04 2.03E-05

Zinc Zn-65 5854.32 BDL --- BDL ---

Cesium Cs-134 18091.1 BDL --- BDL ---




(h)

Short

(wt%) Error

Medium

(wt%) Error

Long

(wt%) Error

Chlorine* Cl-39 0.9367 4.97E-02 1.97E-02 Manganese Mn-56 2.5789 4.65E-02 3.53E-03 Europium Eu-152m 9.3116 6.15E-05 2.79E-06 Potassium K-42 12.355 5.67E-01 2.67E-02 Sodium Na-24 14.997 8.71E-02 1.18E-03 Arsenic As-76 26.24 1.66E-03 1.83E-04



Barium Ba-131 276 5.15E-01 2.41E-03 4.23E-01 2.47E-03

Rubidium Rb-86 447.408 2.86E-03 1.31E-04 2.99E-03 8.79E-05


Chromium Cr-51 664.896 1.02E-03 2.48E-05 1.73E-02 1.81E-04

Iron Fe-59 1067.88 4.10E+00 1.76E-02 5.49E+00 1.26E-02

Mercury Hg-203 1118.26 BDL --- 1.09E-04 9.73E-06


Scandium Sc-46 2010.96 6.41E-04 2.35E-06 6.30E-04 1.31E-06

Selenium Se-75 2874.72 2.13E-05 1.41E-05 BDL ---

Zinc Zn-65 5854.32 7.02E-03 2.20E-04 6.67E-03 1.14E-04

Cesium Cs-134 18091.1 1.27E-04 1.08E-06 1.43E-04 1.98E-06

92




(h)

Short

(wt%) Error

Medium

(wt%) Error

Long

(wt%) Error

Chlorine* Cl-39 0.9367 BDL --- Manganese Mn-56 2.5789 3.55E-02 4.80E-04 Europium Eu-152m 9.3116 9.60E-05 3.68E-06 Potassium K-42 12.355 7.04E-01 1.88E-02 Sodium Na-24 14.997 1.09E-01 8.71E-04 Arsenic As-76 26.24 1.65E-03 9.22E-05



Barium Ba-131 276 1.86E-01 2.53E-03 1.64E-01 2.27E-03

Rubidium Rb-86 447.408 4.23E-03 1.47E-05 3.37E-03 3.42E-05


Chromium Cr-51 664.896 1.05E-02 1.27E-04 5.37E-03 4.07E-05

Iron Fe-59 1067.88 6.80E+00 2.12E-02 7.29E+00 5.83E-03

Mercury Hg-203 1118.26 BDL --- 2.87E-05 4.77E-06


Scandium Sc-46 2010.96 7.21E-04 2.22E-06 7.37E-04 1.27E-06

Selenium Se-75 2874.72 8.59E-05 9.15E-06 1.40E-04 1.01E-05

Zinc Zn-65 5854.32 8.08E-03 2.15E-04 8.41E-03 1.14E-04

Cesium Cs-134 18091.1 1.83E-04 7.88E-06 1.74E-04 1.79E-06

Table D-7. Experimentally determined trace element concentrations of AMD cycle 2



(h)

Short

(wt%) Error

Medium

(wt%) Error

Long

(wt%) Error

Chlorine* Cl-39 0.9367 BDL --- Manganese Mn-56 2.5789 4.37E-03 1.91E-03 Europium Eu-152m 9.3116 BDL --- Potassium K-42 12.355 1.34E-01 2.58E-02 Sodium Na-24 14.997 4.78E+00 7.49E-03 Arsenic As-76 26.24 BDL ---

Lanthanum La-140 40.2852 BDL --- BDL --- 4.40E-04 1.29E-04

Calcium Ca-47 108.864 2.98E+00 2.41E-01 BDL ---

Barium Ba-131 276 3.43E+01 2.35E-02 3.04E+01 1.78E-02

Rubidium Rb-86 447.408 4.08E-04 7.18E-05 5.15E-04 3.91E-05


Chromium Cr-51 664.896 1.43E-02 7.78E-04 BDL ---

Iron Fe-59 1067.88 1.16E+00 1.30E-02 1.16E+00 5.19E-03


Strontium Sr-85 1556.38 4.97E+00 1.28E-02 6.51E+00 7.36E-03

Scandium Sc-46 2010.96 3.35E-06 2.87E-07 4.25E-06 1.18E-07


Zinc Zn-65 5854.32 1.04E-02 2.29E-04 9.83E-03 1.01E-04

Cesium Cs-134 18091.1 BDL --- 1.65E-05 1.28E-06

93




(h)

Short

(wt%) Error

Medium

(wt%) Error

Long

(wt%) Error

Chlorine* Cl-39 0.9367 BDL --- Manganese Mn-56 2.5789 BDL --- Europium Eu-152m 9.3116 BDL --- Potassium K-42 12.355 BDL --- Sodium Na-24 14.997 1.85E+01 2.24E-02 Arsenic As-76 26.24 BDL ---


Calcium Ca-47 108.864 3.58E+00 3.76E-01 Barium Ba-131 276 6.24E+00 1.54E-02 5.23E+00 5.66E-03




Iron Fe-59 1067.88 2.26E-01 9.32E-03 2.11E-01 2.24E-03



Scandium Sc-46 2010.96 6.82E-06 7.05E-07 7.99E-06 1.43E-07


Zinc Zn-65 5854.32 BDL --- 1.27E-04 1.95E-05

Cesium Cs-134 18091.1 BDL --- 2.40E-05 7.35E-07




(h)

Short

(wt%) Error

Medium

(wt%) Error

Long

(wt%) Error

Chlorine* Cl-39 0.9367 BDL --- Manganese Mn-56 2.5789 BDL --- Europium Eu-152m 9.3116 BDL --- Potassium K-42 12.355 BDL --- Sodium Na-24 14.997 1.94E+01 2.32E-02 Arsenic As-76 26.24 BDL ---


Calcium Ca-47 108.864 5.62E+00 6.76E-01

Barium Ba-131 276 9.83E+00 2.35E-02 8.23E+00 7.53E-03

Rubidium Rb-86 447.408 4.82E-04 1.35E-04 4.65E-04 2.99E-05


Chromium Cr-51 664.896 BDL --- 5.44E-04 3.12E-05

Iron Fe-59 1067.88 5.17E-01 1.94E-02 4.52E-01 3.20E-03



Scandium Sc-46 2010.96 4.31E-07 8.76E-07 2.22E-06 9.12E-08


Zinc Zn-65 5854.32 2.76E-04 4.45E-04 5.01E-04 3.08E-05

Cesium Cs-134 18091.1 BDL --- BDL ---

94


01 sample using NAA method


(h)

Short

(wt%) Error

Medium

(wt%) Error

Long

(wt%) Error



Calcium Ca-47 108.864 1.35E+00 8.07E-02 BDL ---

Barium Ba-131 276 5.86E+01 3.29E-02 5.53E+01 2.95E-02

Rubidium Rb-86 447.408 9.13E-04 3.42E-05 8.61E-04 5.17E-05

Protactinium Pa-233 647.4 0.00E+00 0.00E+00 3.45E-03 1.87E-05

Chromium Cr-51 664.896 0.00E+00 0.00E+00 2.82E-02 2.97E-04

Iron Fe-59 1067.88 9.35E-01 9.46E-03 9.52E-01 5.68E-03



Scandium Sc-46 2010.96 9.84E-05 8.49E-07 1.05E-04 4.81E-07

Selenium Se-75 2874.72 0.00E+00 0.00E+00 3.93E-05 3.59E-05

Zinc Zn-65 5854.32 1.85E-02 2.70E-04 1.69E-02 1.33E-04

Cesium Cs-134 18091.1 BDL --- BDL ---


02 sample using NAA method


(h)

Short

(wt%) Error

Medium

(wt%) Error

Long

(wt%) Error



Calcium Ca-47 108.864 1.59E+00 8.95E-02 BDL ---

Barium Ba-131 276 5.32E+01 3.05E-02 5.08E+01 2.78E-02

Rubidium Rb-86 447.408 1.46E-03 9.28E-05 1.14E-03 5.86E-05



Iron Fe-59 1067.88 1.16E+00 4.87E-04 1.19E+00 6.17E-03



Scandium Sc-46 2010.96 1.57E-04 1.07E-06 1.68E-04 6.11E-07

Selenium Se-75 2874.72 BDL --- 6.20E-05 2.89E-05

Zinc Zn-65 5854.32 7.89E-03 2.07E-04 5.94E-03 9.69E-05

Cesium Cs-134 18091.1 BDL --- BDL ---

95

Table D-12. Experimentally determined trace element concentrations of Raw flowhart

solid sample using NAA method


(h)

Short

(wt%) Error

Medium

(wt%) Error

Long

(wt%) Error

Chlorine* Cl-39 0.9367 BDL ---

Manganese Mn-56 2.5789 7.97E-02 4.95E-04

Europium Eu-152m 9.3116 BDL ---

Potassium K-42 12.355 9.23E-01 6.55E-02

Sodium Na-24 14.997 1.29E+00 9.21E-03

Arsenic As-76 26.24 3.93E-03 9.69E-04

Lanthanum La-140 40.2852 1.02E-03 5.81E-06 3.92E-03 3.20E-04

Calcium Ca-47 108.864 3.67E+00 2.12E-01 3.20E+00 2.81E-01

Barium Ba-131 276 3.33E+00 6.72E-03 3.17E+00 4.97E-03

Rubidium Rb-86 447.408 4.60E-03 1.55E-04 5.29E-03 1.01E-04


Chromium Cr-51 664.896 5.87E-03 1.38E-04 1.12E-02 8.71E-05

Iron Fe-59 1067.88 5.56E+00 2.52E-02 5.67E+00 1.33E-02

Mercury Hg-203 1118.26 BDL --- 6.73E-05 1.41E-05


Scandium Sc-46 2010.96 4.50E-04 1.96E-06 4.79E-04 1.15E-06


Zinc Zn-65 5854.32 1.35E-02 6.30E-04 BDL ---

Cesium Cs-134 18091.1 1.75E-03 1.40E-05 1.90E-03 7.99E-06

Table D-13. Experimentally determined trace element concentrations of FS3 Effluent

sample using NAA method


(h)

Short

(wt%) Error

Medium

(wt%) Error

Long

(wt%) Error

Chlorine* Cl-39 0.9367 1.34E-03 1.62E-02

Manganese Mn-56 2.5789 1.84E-01 8.87E-04


Potassium K-42 12.355 6.35E-01 7.45E-02

Sodium Na-24 14.997 1.72E+00 1.16E-02

Arsenic As-76 26.24 BDL ---


Calcium Ca-47 108.864 7.74E+00 2.80E-01 6.90E+00 1.55E-01

Barium Ba-131 276 5.87E+00 8.53E-03 5.23E+00 5.96E-03

Rubidium Rb-86 447.408 3.76E-03 1.30E-04 3.67E-03 7.44E-05


Chromium Cr-51 664.896 6.03E-03 1.38E-04 5.36E-04 2.20E-05

Iron Fe-59 1067.88 1.61E+00 1.55E-02 1.59E+00 6.47E-03



Scandium Sc-46 2010.96 4.39E-04 1.99E-06 4.71E-04 1.18E-06


Zinc Zn-65 5854.32 7.46E-03 5.51E-04 7.20E-03 1.08E-04

Cesium Cs-134 18091.1 2.14E-04 8.93E-06 1.77E-04 1.87E-06

96

Table D-14. Experimentally determined trace element concentrations of Marcellus

Flowback sample using NAA method


(h)

Short

(wt%) Error

Medium

(wt%) Error

Long

(wt%) Error


Manganese Mn-56 2.5789 2.15E-03 3.74E-04


Potassium K-42 12.355 2.46E-02 9.67E-02

Sodium Na-24 14.997 1.46E+01 4.66E-02

Arsenic As-76 26.24 5.79E-04 3.88E-03

Lanthanum La-140 40.2852 BDL --- BDL ---

Calcium Ca-47 108.864 3.79E+00 2.30E-01 3.50E+00 1.54E-01

Barium Ba-131 276 2.15E+00 4.28E-03 1.93E+00 2.94E-03

Rubidium Rb-86 447.408 3.68E-04 1.44E-04 5.11E-04 3.41E-05

Protactinium Pa-233 647.4 1.94E-05 6.59E-06 BDL ---


Iron Fe-59 1067.88 8.50E-02 6.03E-03 8.64E-02 1.37E-03



Scandium Sc-46 2010.96 1.50E-06 3.39E-07 1.19E-06 6.53E-08


Zinc Zn-65 5854.32 N/A N/A 4.62E-04 2.90E-05

Cesium Cs-134 18091.1 1.06E-04 3.79E-06 1.22E-04 1.15E-06

Table D-15. Experimentally determined trace element concentrations Franklin discharge



(h)

Short

(wt%) Error

Medium

(wt%) Error

Long

(wt%) Error


Manganese Mn-56 2.5789 BDL ---


Potassium K-42 12.355 BDL ---

Sodium Na-24 14.997 1.23E+01 5.21E-02



Calcium Ca-47 108.864 8.00E+00 2.74E-01 5.65E+00 1.36E-01

Barium Ba-131 276 7.70E-03 3.69E-03 4.01E-03 2.66E-04




Iron Fe-59 1067.88 BDL --- BDL ---



Scandium Sc-46 2010.96 BDL --- BDL ---


Zinc Zn-65 5854.32 BDL --- BDL ---

Cesium Cs-134 18091.1 BDL --- 6.78E-06 5.06E-07

97

Table D-16. Experimentally determined trace element concentrations Sample 01 liquid



(h)

Short

(wt%) Error

Medium

(wt%) Error

Long

(wt%) Error

Chlorine* Cl-39 0.9367 1.36E-02 8.99E-02

Manganese Mn-56 2.5789 4.06E-03 6.59E-04



Sodium Na-24 14.997 1.86E+01 5.58E-01



Calcium Ca-47 108.864 5.60E+00 2.63E-01 4.25E+00 1.70E-01

Barium Ba-131 276 4.69E-01 2.34E-02 4.17E-01 1.48E-02

Rubidium Rb-86 447.408 4.63E-04 1.55E-04 7.30E-04 2.98E-05



Iron Fe-59 1067.88 3.53E-02 7.09E-03 4.09E-02 1.08E-03





Zinc Zn-65 5854.32 BDL --- BDL ---

Cesium Cs-134 18091.1 8.94E-05 2.13E-06 9.76E-05 1.07E-06




(h)

Short

(wt%) Error

Medium

(wt%) Error

Long

(wt%) Error


Manganese Mn-56 2.5789 5.51E-03 6.85E-04



Sodium Na-24 14.997 1.66E+01 8.62E-01



Calcium Ca-47 108.864 5.64E+00 2.81E-01 3.96E+00 1.36E-01

Barium Ba-131 276 5.12E-01 2.76E-02 4.05E-01 1.53E-02

Rubidium Rb-86 447.408 5.96E-04 8.02E-05 6.92E-04 2.34E-05



Iron Fe-59 1067.88 3.46E-02 2.47E-03 3.91E-02 1.19E-03





Zinc Zn-65 5854.32 BDL --- BDL ---

Cesium Cs-134 18091.1 7.65E-05 2.01E-06 8.66E-05 1.10E-06

98




(h)

Short

(wt%) Error

Medium

(wt%) Error

Long

(wt%) Error


Manganese Mn-56 2.5789 1.87E-02 2.35E-03


Potassium K-42 12.355 8.01E-01 4.17E-02

Sodium Na-24 14.997 1.14E+01 6.56E-01



Calcium Ca-47 108.864 1.04E+01 7.02E-01 5.39E+00 1.08E-01

Barium Ba-131 276 2.60E-03 3.29E-04 3.27E-03 2.76E-04

Rubidium Rb-86 447.408 9.98E-04 9.16E-05 1.22E-03 2.06E-04



Iron Fe-59 1067.88 6.00E-02 4.78E-03 5.47E-02 1.11E-03


Strontium Sr-85 1556.38 2.43E-01 1.91E-02 1.22E+00 3.95E-02

Scandium Sc-46 2010.96 BDL --- 4.05E-08 3.26E-08


Zinc Zn-65 5854.32 5.64E-04 2.78E-04 6.44E-04 2.97E-05

Cesium Cs-134 18091.1 4.33E-05 1.50E-06 4.30E-05 6.38E-07




(h)

Short

(wt%) Error

Medium

(wt%) Error

Long

(wt%) Error

Chlorine* Cl-39 0.9367 1.12E+00 9.09E-01

Manganese Mn-56 2.5789 2.35E-03 1.18E-03


Potassium K-42 12.355 4.17E-01 1.32E-01

Sodium Na-24 14.997 2.86E+00 3.92E-02



Calcium Ca-47 108.864 1.93E+00 7.08E-01 1.16E+00 2.32E-01

Barium Ba-131 276 3.25E-03 2.58E-03 5.24E-03 2.34E-03

Rubidium Rb-86 447.408 BDL --- BDL ---



Iron Fe-59 1067.88 2.99E-02 9.75E-03 3.30E-02 4.30E-03

Mercury Hg-203 1118.26 BDL --- 1.26E-04 3.14E-05

Strontium Sr-85 1556.38 BDL --- 2.65E-02 7.36E-03

Scandium Sc-46 2010.96 3.99E-06 8.57E-07 3.59E-06 4.51E-07


Zinc Zn-65 5854.32 3.06E-03 5.64E-04 3.27E-03 2.86E-04

Cesium Cs-134 18091.1 BDL --- BDL ---

99




(h)

Short

(wt%) Error

Medium

(wt%) Error

Long

(wt%) Error


Manganese Mn-56 2.5789 1.71E-03 6.32E-04



Sodium Na-24 14.997 3.32E+00 3.18E-02


Lanthanum La-140 40.2852 6.19E-06 1.49E-08 BDL ---

Calcium Ca-47 108.864 4.48E+00 7.02E-01 1.82E+00 2.38E-01

Barium Ba-131 276 4.85E-02 1.74E-03 5.04E-02 1.53E-03




Iron Fe-59 1067.88 BDL --- 2.06E-03 9.00E-04



Scandium Sc-46 2010.96 4.86E-07 1.07E-06 1.64E-06 2.39E-07


Zinc Zn-65 5854.32 BDL --- BDL ---

Cesium Cs-134 18091.1 1.34E-05 2.55E-06 1.15E-05 1.30E-06

Table D-21. Experimentally determined trace element concentrations HR, Evop Test 03



(h)

Short

(wt%) Error

Medium

(wt%) Error

Long

(wt%) Error


Manganese Mn-56 2.5789 7.55E-02 5.62E-04



Sodium Na-24 14.997 8.09E-01 6.57E-03



Calcium Ca-47 108.864 6.83E-01 6.10E-02 6.00E-01 4.65E-02

Barium Ba-131 276 4.27E+01 2.62E-02 4.46E+01 2.40E-02

Rubidium Rb-86 447.408 5.39E-04 5.51E-05 4.72E-04 2.48E-05

Protactinium Pa-233 647.4 3.17E-03 2.51E-05 BDL ---

Chromium Cr-51 664.896 2.31E-02 1.51E-03 BDL ---

Iron Fe-59 1067.88 1.69E-02 1.93E-03 1.50E-02 9.73E-04



Scandium Sc-46 2010.96 BDL --- 2.39E-07 5.03E-08


Zinc Zn-65 5854.32 2.24E-03 2.04E-04 4.89E-04 3.40E-05

Cesium Cs-134 18091.1 BDL --- BDL ---

The trace element concentrations determined by NAA method and interlaboratory study

(MPV) are compared and graphically demonstrated in Figures D-1 through D-6.

100

Figure D-1. A graphical comparison of sodium concentrations in oil and gas wastewater

samples.

Figure D-2. A graphical comparison of potassium concentrations in oil and gas wastewater

samples.

27000 24760.8

47500

39596.5

66850

31271.4

0.0E+00

1.0E+04

2.0E+04

3.0E+04

4.0E+04

5.0E+04

6.0E+04

7.0E+04

8.0E+04

Co

nce

ntr

atio

n (

mg/l

)


Sodium

Most probable value (MPV) NAA measured value

336

B.D.L.

716

B.D.L.

2190 2203.0

0.0E+00

5.0E+02

1.0E+03

1.5E+03

2.0E+03

2.5E+03

Conce

ntr

atio

n (

mg/l

)


Potassium


101

Figure D-3. A graphical comparison of calcium concentrations in oil and gas wastewater

samples.

Figure D-4. A graphical comparison of strontium concentrations in oil and gas wastewater

samples.

10000

7434.7

19800

13471.8

31400

28543.3

0.0E+00

5.0E+03

1.0E+04

1.5E+04

2.0E+04

2.5E+04

3.0E+04

3.5E+04C

on

cen

trat

ion

(m

g/l

)


Calcium


2160 2162.9

3710

4083.8

1540

669.3

0.0E+00

5.0E+02

1.0E+03

1.5E+03

2.0E+03

2.5E+03

3.0E+03

3.5E+03

4.0E+03

4.5E+03

Conce

ntr

atio

n (

mg/l

)


Strontium


102

Figure D-5. A graphical comparison of barium concentrations in oil and gas wastewater

samples.

Figure D-6. A graphical comparison of iron concentrations in oil and gas wastewater

samples.

659 622.4

13201221.8

6.12 7.20.0E+00

2.0E+02

4.0E+02

6.0E+02

8.0E+02

1.0E+03

1.2E+03

1.4E+03

1.6E+03

Co

nce

ntr

atio

n (

mg/l

)


Barium


64.854.3

94.982.8

169 165.1

0.0E+00

2.0E+01

4.0E+01

6.0E+01

8.0E+01

1.0E+02

1.2E+02

1.4E+02

1.6E+02

1.8E+02

2.0E+02

Co

nce

ntr

atio

n (

mg/l

)


Iron


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