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DOE Review Release 01/12/2017

Locations of Stable Metal Tritide Use at the 

Savannah River Site

ORAUT-RPRT-0072 Rev. 00

Effective Date: 01/09/2017

Supersedes: None

Subject Expert(s): Jason Davis

Document Owner 

Approval: Signature on File

James M. Mahathy, Document Owner

Approval Date: 01/04/2017

Concurrence: Signature on File

Daniel H. Stempfley, Objective 4 Representative

Concurrence Date: 01/05/2017

Concurrence: Vickie S. Short Signature on File for

Kate Kimpan, Project Director

Concurrence Date: 01/04/2017

Approval: Signature on File

James W. Neton, Associate Director for Science

Approval Date: 01/09/2017

FOR DOCUMENTS MARKED AS A TOTAL REWRITE, REVISION, OR PAGE CHANGE, REPLACE THE PRIOR 

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.

New Total Rewrite Revision Page Change
Document No. ORAUT-RPRT-0072 Revision No. 00 Effective Date: 01/09/2017 Page 2 of 31

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

01/09/2017 00 New document initiated in response to questions raised by SC&A 

that special tritium compounds have not been considered in the 

analysis of personnel exposures at the Savannah River Site. 

Incorporates formal internal and NIOSH review comments. Training 

required: As determined by the Objective Manager. Initiated by 

Jason Davis.

TRADEMARK INFORMATION

ZircaloyTM is a trademark of Westinghouse Electric Corporation in the United States and/or other 

countries.

All other trademarks are the property of their respective owners.
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TABLE OF CONTENTS

SECTION TITLE PAGE

Acronyms and Abbreviations ............................................................................................................... 4

1.0 Summary ................................................................................................................................. 6

2.0 Introduction .............................................................................................................................. 6

2.1 General Properties of Tritium........................................................................................ 6

2.2 Stable Metal Tritides..................................................................................................... 7

3.0 Use of Metal Tritides ................................................................................................................ 8

3.1 Zeolite .......................................................................................................................... 9

3.2 Stainless Steel.............................................................................................................. 9

3.3 Uranium...................................................................................................................... 10

3.4 Magnesium................................................................................................................. 11

3.5 Lanthanum-Nickel-Aluminum Alloy ............................................................................. 12

3.6 Titanium ..................................................................................................................... 13

3.7 Palladium ................................................................................................................... 13

3.8 Intentionally Formed Zirconium................................................................................... 15

3.9 Zirconium Tritides from Fuels ..................................................................................... 15

4.0 Stable Metal Tritide Handling Areas ....................................................................................... 17

4.1 F-Area ........................................................................................................................ 17

4.2 H-Area........................................................................................................................ 18

4.3 A-Area........................................................................................................................ 19

5.0 Potential for Metal Hydride Exposure ..................................................................................... 21

5.1 Engineering Controls .................................................................................................. 21

5.2 Shortfalls of Existing Controls..................................................................................... 22

6.0 Dosimetry Considerations ...................................................................................................... 23

6.1 Solubility..................................................................................................................... 23

6.2 Sample Self-Absorption.............................................................................................. 25

6.3 Determination of Intakes from Existing Data............................................................... 25

6.3.1 Air Sampling Data ........................................................................................ 25

6.3.2 Surface Contamination Data ........................................................................ 26

7.0 Conclusion ............................................................................................................................. 26

References ........................................................................................................................................ 27

LIST OF TABLES

TABLE TITLE PAGE

6-1 Absorption types of representative metal tritides .................................................................... 25
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ACRONYMS AND ABBREVIATIONS 

CFR Code of Federal Regulations

Ci curie

cm centimeter

DOE U.S. Department of Energy

dpm disintegrations per minute

g gram

GI gastrointestinal

HSV hydride storage vessel

HT elemental tritium (tritiated gas)

HTO tritiated water vapor

HTV hydrogen transport vessel

in. inch 

kBq kilobequerel

keV kiloelectron-volt (1,000 electron-volts)

kg kilogram

L liter

LANA lanthanum-nickel-aluminum alloy (LaNi4.25Al0.75 only) 

LSC liquid scintillation counting

LWR light-water reactor

mol mole

MPa megapascal

MPD multistage palladium diffuser

NIOSH National Institute for Occupational Safety and Health

ORAU Oak Ridge Associated Universities

ppm parts per million

psia pounds per square inch absolute

PUREX plutonium-uranium extraction

PWR pressurized-water reactor

RSI Resource Services Incorporated

R&D research and development

SAF self-absorption factor

SMT stable metal tritide

SRDB Ref ID Site Research Database Reference Identification (number)

SRL Savannah River Laboratory

SRNL Savannah River National Laboratory

SRP Savannah River Plant

SRS Savannah River Site
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SRTC Savannah River Technology Center

STC special tritium compound

SUF serum ultrafiltrate

TCAP thermal cycling absorption process

TPBAR tritium-producing burnable absorber rod

TVA Tennessee Valley Authority

U.S.C. United States Code

UREX uranium extraction

wt% weight percent

μCi microcurie

μm micrometer

°C degrees Celsius
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1.0 SUMMARY

NOTE: Because of the sensitive nature of some of the processes at the U.S. Department of 

Energy (DOE) Savannah River Site (SRS), this report does not provide a comprehensive list of 

metal tritides SRS has used. Rather, this report provides a representation of the types and 

uses of metal tritides that might have been encountered at the site. Evaluation of other metal 

hydrides is conducted in a set of classified notes associated with this document.

Various metal tritides are encountered at SRS Tritium Facilities. In the early 1980s, the Savannah 

River Laboratory began a major program to develop and use metal hydrides in its tritium production 

facilities. Metal hydrides turned out to be ideally suited for tritium handling and processing. These 

metals include uranium, palladium, titanium, lanthanum-aluminum-nickel tritide, mischmetal, and 

zirconium. Evaluations of the physical and chemical properties of these compounds have shown that 

uranium and palladium do not bind tritium in a manner significantly different from tritium oxide.

Therefore, they are not considered as special tritium compounds (STCs). The remaining metals form 

compounds that cannot be adequately assessed using the tritium oxide urine bioassay model and are 

therefore classified as STCs.

2.0 INTRODUCTION

At SRS, several metals have been used as a means of separating and storing tritium. These metals 

include uranium, palladium, titanium, lanthanum-aluminum-nickel tritide, lithium tritide, tritiated 

mercury, mischmetal, and zirconium. Tritium can form metal tritides with the relative ease of tritium 

exchange. Metal tritides have been used and researched at SRS, other DOE sites, and commercial 

sites for many years.

These compounds can be created by intentional combination of tritium with the desired materials or 

by inadvertent contamination of a material that has been subjected to the presence of tritium over

time. Metal tritides are essentially tritium in a particulate form, which can present dose reconstruction 

problems that are distinctly different from those of the more commonly encountered tritiated water 

vapor (HTO).

The primary objective of this document is to provide information on the potential sources of stable 

tritium compounds at SRS. The source documents that were reviewed for this report consist of those 

in the Site Research Database (SRDB) as well as documents available in the open literature. This 

document reports information about tritium compounds that has been found by the document review 

process to date only. 

2.1 GENERAL PROPERTIES OF TRITIUM

Tritium undergoes decay by emission of a low-energy beta particle and has a half-life of 

approximately 12.3 years. Due to their low penetrating ability, tritium emissions are not readily 

detectable through the use of instruments that are commonly used to monitor for other common 

radionuclides that can contaminate the workplace (e.g., pancake Geiger-Müller detectors connected 

to count rate meters).

The most common technique used for tritium air sampling (for both HT and HTO) is the flow-through 

ionization chamber. Removable tritium surface contamination can be detected through liquid 

scintillation counting (LSC), and this analytical method is in widespread use for analysis of workplace 

(contamination smears) and biological (urine) sample media.

Because it decays by emission of low-energy beta particles that cannot penetrate human skin, tritium 

is not typically considered a significant external radiation exposure hazard. However, inhalation, 
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ingestion, or skin absorption of tritium can result in internal radiation doses. Because tritium in its 

prevalent oxide form behaves chemically identically to water, it undergoes fairly rapid uptake by, and 

elimination from, the human body. These biological processes make tritium oxide relatively easy to 

detect in bodily excretions (primarily urine) at very low levels, which simplifies the radiobioassay, dose 

assessment, and exposure control processes.

Hydrogen atoms (including those that exist in the form of tritium) tend to diffuse through materials with 

which they come into contact and can be released at a later time. While this property of hydrogen 

atoms can complicate measures to contain tritium, it has been well characterized and appropriate 

materials can be specified to minimize unintended adverse effects. This property can also be used to 

practical advantage because hydrogen atoms can be diffused through other materials and stored 

within the matrix for subsequent storage, transportation, release, and use. The diffusion processes 

can be accelerated through application of heat.

2.2 STABLE METAL TRITIDES 

Other than HT and HTO, tritium can form metal tritides with the relative ease of tritium exchange. The 

property of interest of these materials is that they react with hydrogen isotopes to form stochiometric 

hydrides, where the hydrogen isotope is contained in interstitial sites in the metal lattice, rather than 

occupying binding sites of the metal itself (Nobile and Motyka 1991).

The hydriding/dehydriding reaction of various metals can be used as a basis for the storage, pumping, 

compression, purification, and separation of hydrogen isotopes. One advantage of using hydride 

technology in tritium processes is the compact size of hydride equipment. This is due to the fact that 

the volumetric density of hydrogen in typical hydrides is greater even than that of liquid or solid 

hydrogen. The compact size of hydride beds allows installation in process hoods and gloveboxes at 

locations where only minimal space is available. For example, use of these materials at SRS has 

allowed for the reduction of the number of gloveboxes in use and the overall size of some of the 

facilities. Other benefits of using hydride technology for tritium handling applications are improved 

safety during storage of tritium in the hydride form at below atmospheric pressure and simple, reliable 

pumps and compressors with valves as the only moving parts (Ortman et al. 1989).

Metal tritides are classified as either stable or unstable depending on their ease to release tritium. 

The unstable metal tritides pose no problems because they dissociate themselves with ease in the 

environment and biologically form the more conventional water vapor tritium. Stable metal tritides 

(SMTs), however, maintain their original form with a stable metal-tritium bond. Traditionally, SMTs 

have not been well understood in terms of biokinetlcs, bioassay, workplace monitoring techniques, 

and characterization and identification in the workplace.

Particulate SMTs can be ingested or inhaled and result in radiation dose. After ingestion, these 

particles lose some fraction of their tritium as HTO in stomach digestive fluids. Some SMTs lose 

essentially all of the tritium, and others lose very little, depending on the solubility of the specific 

material in the gastrointestinal (GI) tract. In any case, the only significant dose from ingestion of 

SMTs is due to the resulting HTO that is released. Because HTO is rapidly assimilated 

physiologically, this dose component is readily assessed via urine bioassay. After inhalation, 

however, particulate SMTs can be deposited in the lung. Once deposited, the tritium bound to the 

particulate is removed from the body via two mechanisms:

1. Dissolution of tritium from the particulate and absorption into the body as HTO and removal 

through urine, expired air, and perspiration; and
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2. Mechanical transport of the particulate itself to the GI tract and removal through feces. 

(Inhaled SMTs removed to the gut will cause dose in the same manner as ingested SMTs, as 

described above.) 

The behavior of SMTs can present a variety of challenges to the reconstruction of doses based on 

records from a facility’s radiation protection program. For example:

• It can be difficult to differentiate between SMTs and the more common forms of tritium using 

routine workplace monitoring techniques.

• The physical and chemical behavior of SMTs can render commonly used tritium bioassay and 

available internal dose models ineffective.

• Difficulties in identifying and quantifying contamination can cause significant delays between 

performance of workplace monitoring and completion of analyses.

Although the DOE radiological protection community has been aware of SMTs for many years, to date 

their impact has been limited by the design features that are incorporated into DOE facilities that 

handle significant quantities of tritium. These design features include various forms of material 

containment and control, such as gloveboxes and high-efficiency particulate air filtering ventilation 

systems that effectively prevent significant releases of SMTs to occupied areas of the workplace or 

the environment. In recent years, sensitivity to SMT contamination has increased as a result of DOE 

decontamination and decommissioning of older facilities. These activities can compromise the 

effectiveness of the installed design features and allow releases of SMT contamination to the 

surrounding areas.

NIOSH has established a method for assigning doses due to intakes of STCs. This process provides 

recommendations for the calculation of doses based tritium bioassay. These recommendations vary 

depending on what is known about the STC that is contributing to the urinary excretion of tritium 

(ORAUT 2007). 

3.0 USE OF METAL TRITIDES 

NOTE: Because of the sensitive nature of some of the processes at SRS, this report does not 

provide a comprehensive list of metal tritides SRS has used. Rather, this report provides a 

representation of the types and uses of metal tritides that might have been encountered at the 

site. Evaluation of other metal hydrides is conducted in a set of classified notes associated 

with this document.

The SRS Tritium Facilities have used a variety of metal hydride materials in process vessels or beds 

for the safe processing and storage of tritium. Due to the sensitive nature of the work at SRS, 

information such as the quantities of material and the details of the work on metal hydrides and tritides 

at SRS are still classified. However, a qualitative breakdown of the metal hydride work is presented 

here. At SRS, these materials are used in tritium processes as storage media and are the subject (or 

byproduct) of research activities (Moxley 2002).

At SRS, the determination as to whether a compound is classified as an STC is based on the degree 

to which the tritium atoms are held in the compound. The tritium in STCs is bound more tightly than 

the tritium in tritium oxide. Using this designation, uranium tritide and palladium tritide are not 

considered SMTs because they do not bind tritium in a manner significantly different from tritium oxide 

(Moxley 2002). To develop a complete picture of the use and production of SMTs at the site, this 

report includes discussion of these two materials.
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A program was initiated in 1981 by the Savannah River Laboratory in cooperation with the Tritium 

Facilities with the goal of implementing metal hydride technology in the SRS Tritium Facilities. This 

program was initiated to take advantage of the discovery in 1969 of intermetallic compounds, such as 

LaNi5 and FeTi, that would rapidly and reversibly absorb and desorb hydrogen under moderate 

conditions of temperature and pressure. Since then the processes described below have gone 

through bench-scale, pilot-scale and, in some cases, full-scale testing.

3.1 ZEOLITE

For approximately 64 years, SRS has isolated tritium from its process streams by catalyzing the 

formation of tritiated water (from process streams) and then sorbing that water on a 3A zeolite

molecular sieve (molsieve) bed. This requires converting elemental tritium to tritium oxide and 

capturing the oxide on zeolite moisture absorber rods, which can then be processed to recover tritium. 

The tritium is recovered by regenerating the saturated bed into a magnesium-based water-cracking 

unit (Hsu and Heung 1997).

The SRS Tritium Facilities use 30 nitrogen gloveboxes and 2 primary stripper systems that support 

normal tritium removal from the nitrogen atmosphere. A third identical system named the secondary 

stripper handles accidental tritium release in any of the gloveboxes. Each of these three systems has 

three zeolite beds, with each of these beds holding 104 kg of zeolite type 3A. A fourth stripper system 

named the purge stripper, which controls the purge of nitrogen from the gloveboxes to the 

atmosphere, has three smaller zeolite beds. Each of these smaller beds holds 27 kg of zeolite 

type 4A (Heung et al. 2004).

Experience from the stripper system at SRS shows that more than 99.9% of the moisture on zeolite 

beds is nontritiated water. About half of the water appears to come from moisture leaking into 

gloveboxes, with about half due to negative pressure operation and half to glove permeation (Hsu and 

Heung 1997).

Because zeolite molsieves rely on strong physical forces rather than chemisorption to retain 

adsorbates, their adsorption is characterized by a non-linear isotherm (the amount of a given 

compound adsorbed increases rapidly to a saturation value as its pressure or concentration increases 

in the external bulk phase). Any further increase in pressure at constant temperature causes no 

further increase in the amount adsorbed. With zeolite molsieves, this equilibrium saturation value 

typically corresponds to a complete filling of the internal void volume with the adsorbate. When 

adsorbed molecules are desorbed via heat or by displacement with another material, the crystal’s 

chemical state remains unchanged (EPA 1999).

Because the tritium is adsorbed on the zeolite rather than forming an alloy with the metallic 

components, zeolite is not considered a metal tritide in this report.

3.2 STAINLESS STEEL

Austenitic stainless steels are extensively used in the nuclear industry because of their good corrosion 

resistance and mechanical properties. Prolonged use of tritium in process plants results in a tritium 

burden in steel and other metals (Richardson and Hong 2001). A typical 1020 steel component 

exposed to hydrogen gas at 1 atmosphere of pressure at room temperature will dissolve far less than 

1 ppm of hydrogen. However, when the same component is corroded by acid cleaning, exposure to 

hydrogen sulfide, or simple exposure to moisture, the concentration of nascent hydrogen on the 

surface can become quite high and the hydrogen solubility during the exposure can exceed several 

parts per million, but might not lead to the formation of a hydride (Louthan 2008).
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The equilibrium solubility of hydrogen in iron at a hydrogen pressure of 1 bar is small. Theoretical 

concentrations of interstitial hydrogen in solid iron vary from an atomic ratio of about 1 × 10-6 at room 

temperature to approximately 1 × 10-3 near the melting point of iron. The preparation of bulk, 

composition, crystal structure, and physical properties, however, is possible only under equilibrium 

conditions in an atmosphere of hydrogen at pressures of several gigapascals (Antonov et al. 2002).

At first glance this solubility determination contradicts the observed behavior of tritium retention in 

steel. However real solids are not perfect crystals, but rather they contain defects that can play a 

crucial role in the uptake and transport of hydrogen in metallic materials. The absorbed hydrogen in a 

metal can be present in either of two locations in the microstructure: interstitial sites in the lattice or 

extraordinary sites typically associated with crystalline defects. Hydrogen absorption causes the iron 

lattice to expand because the effective size of the nascent hydrogen atom is larger than the size of the 

interstitial site. This size difference causes hydrogen to prefer extraordinary sites where the site size 

within the lattice is expanded and the incoming hydrogen is more readily accommodated. Examples 

of extraordinary sites include grain boundaries, vacancies, dislocations, and any other volume where 

the sites within the lattice structure are dilated.

The concentration of hydrogen at the extraordinary sites in body-centered cubic (ferritic) or 

body-centered tetragonal (martensitic) steels can greatly exceed the concentration of hydrogen at the 

normal inter-lattice sites. Frequently, the hydrogen at extraordinary sites is termed “trapped 

hydrogen,” and the hydrogen at the normal inter-lattice sites is termed “dissolved or diffusible

hydrogen.” The mobility of hydrogen increases as temperature increases, and the difference in 

residence time at the various sites eventually becomes negligible at high temperatures. Further, at 

temperatures much below room temperature the mobility of hydrogen decreases to the point that 

relocation cannot occur and local equilibrium cannot be maintained (Louthan 2008). 

Iron and steel hydrides are not considered further in this report because the high temperatures and 

pressures that are required for formation make it very unlikely that these materials would be 

encountered at the SRS. 

3.3 URANIUM

As previously mentioned, metal tritides can be created either by the intentional combination of tritium 

with the desired materials or by inadvertent contamination of a material that has been subjected to the 

presence of tritium over time. Before 1981, metal hydrides were used to process tritium at DOE sites, 

such as uranium storage and purification beds at Mound Laboratory and Los Alamos National 

Laboratory, a vanadium compressor bed at Los Alamos National Laboratory, and uranium beds at the 

Savannah River Laboratory for a short time in the 1950s to store and transport deuterium and tritium 

during experimental work.

Depleted uranium has a long-standing history in tritium activities. This metal is used to store tritium, 

transport tritium gas and, in special applications, selectively remove tritium from flowing gas streams. 

Uranium offers a very high storage capacity; 120 standard cm3 of gas per gram of uranium can be 

easily accommodated without compromising the reaction kinetics. It is well suited to removing tritium 

from clean inert gas streams with high tritium concentrations and offers uniform reaction kinetics over 

a very broad range of tritium-to-uranium atom ratios. It is possible to charge uranium with tritium up to 

an atom ratio of nearly 3 (tritium atoms per uranium atom), although the operating range is usually 

restricted to a ratio of 1.5 so sufficient tritium storage capacity remains for unforeseen events 

(Shmayda 2005).

At room temperature, tritium in the presence of uranium powder forms uranium tritide. The tritium 

partial pressure in the bed is very low. As a result, at room temperature the bed acts as a vacuum 

pump that getters all of the hydrogen isotopes. The impurity gases that might be present, such as 
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3

He, N2, O2, or argon, either remain in the overpressure gas in the bed or react with uranium to form 

stable compounds.

Uranium has a theoretical maximum storage capacity of 1.26 wt% and bonds with tritium to form a 

tritide at room temperature and pressure. However, uranium as a storage material has drawbacks for 

various operations. Tritide formation in uranium is an exothermic reaction meaning the system needs 

to be actively cooled to maintain the fast kinetics of the tritiding process. There is also a significant 

volume expansion as the reaction takes place, which limits the size and orientation of storage beds. 

At SRS, high temperatures were required to desorb the tritium from the metal, which led to concerns 

about tritium permeation through the stainless-steel bed and process inefficiencies due to long 

heating and cooling times. In addition, the pyrophoric nature of uranium led to safety concerns with 

respect to its potential use at SRS (Ortman et al. 1989).

In the fourth quarter of 1991, a program to develop a hydrogen transport vessel (HTV) using depleted 

uranium with capacity for up to 18 g of tritium was initiated. An experimental transport vessel was 

fabricated and tested in the laboratory before the final design was completed (Heung 1994a). The 

vessel body is fabricated from two 4-in. schedule 40 stainless-steel pipe caps. This vessel contains 

0.5 kg of depleted uranium in the form of 0.5- to 1.0-in.-diameter bar stock in lengths not greater than 

0.5 in. Before first use, the depleted uranium in a new HTV must be activated. During activation, 

hydrogen gas reacts with the uranium bar stock to decompose the bar stock into powder. Heating 

during activation accelerates penetration of any uranium oxide layer by the hydrogen isotopes. Port 

filters with a maximum pore size of 2 μm were designed to contain all uranium powder within the HTV. 

This hydride transportation bed has been in service since 1995 (Heung 2001).

SRS has also used uranium as a decomposer to crack tritiated water from various streams into 

oxygen and hydrogen isotopes. These isotopes can then be separated from any helium isotopes 

present by permeation through a palladium-silver alloy diffuser (Ortman et al. 1989). Because the 

decomposer is operated at high temperatures, it is unlikely that the exothermic hydriding reaction 

would occur to an appreciable degree as a result of this process.

3.4 MAGNESIUM

The regeneration and recovery of hydrogen isotopes from zeolite beds is accomplished by a zeolite 

bed recovery system. The system consists of magnesium beds, a furnace, circulation pumps, 

pressure regulators, mass flow meters, and recycle tanks. The magnesium bed is a 25-cm diameter 

vessel and holds 20 kg (823 mol) magnesium turnings supported by a porous plate at the bottom. 

The magnesium container is heated in a furnace equipped with resistant electric heaters. Inlet gas 

containing water moisture flows through a standpipe to the lower side of the porous plate. As the gas 

mixture rises through the magnesium bed, the moisture reacts with the magnesium to irreversibly form 

MgO and elemental hydrogen. The hydrogen gas exits the top of the bed and circulates in the loop. 

During operation the magnesium bed is maintained between 450°C and 500°C. Twenty kilograms of 

magnesium can reduce a maximum of 823 mol of water and generate 18,000 L of H2 (Heung et al. 

2004). 

When the magnesium is consumed, the bed is removed and disposed of as low-level radioactive 

waste. The hydrogen isotopes are collected in recycle tanks and later separated to recover the tritium 

(Heung et al. 2004). It is unlikely that the magnesium would have been able to form a hydride, 

however. The hydrogenation/dehydrogenation reaction for pure magnesium is very slow and occurs 

only at high temperature (350°C to 400°C) under a hydrogen pressure of more than 3 MPa. Further, 

the surface oxidation of magnesium (MgO) exposed to air would significantly hinder the rate of 

hydrogenation (Stier et al. 2005).
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3.5 LANTHANUM-NICKEL-ALUMINUM ALLOY

The first metal hydride vessel designed for tritium service at SRS was a 6-mol capacity 

storage/compressor. It started tritium storage in 1984 and was removed from service in 1986. The 

vessel was used to store tritium at ambient room temperature and pressure and to produce up to 

300 psia by heat. The container volume was filled with 1 kg of lanthanum-nickel-aluminum alloy 

(LANA; specifically LaNi4.7Al0.3 in this instance), which occupied about 75% of the volume. The 

remaining 25% of volume was left for volume expansion of the hydride particles. Inside the inlet/outlet

port of the vessel was a cylindrical, sintered, porous stainless-steel filter with a nominal pore size of 

5 μm (Heung 2001).

An experimental tritium gas handling manifold was installed in 1985 in support of the development of 

metal hydride technology. The manifold uses a small LaNi4.25Al0.75 metal hydride storage bed to store 

about 10 g of tritium for experimental work. As tritium is needed in the manifold, this storage bed is 

heated to deliver it at the required pressures. When tritium is no longer needed, it is reabsorbed by 

cooling the storage bed. This minimizes the amount of tritium that must be pumped through 

mechanical vacuum pumps, which prolongs the life of the pumps (Ortman et al. 1989). 

Twenty LANA hydride samples in a controlled matrix study were charged with tritium and placed in 

storage in March and April 1987. The interaction of tritium with these samples was investigated by 

collection of equilibrium desorption isotherms at time intervals generally on the order of several 

months (Ortman et al. 1989).

After the success of the 6-mol beds, larger beds with capacities up to 70 mol were designed for 

isotope storage and pumping. These beds are used to store tritium and its isotope mixtures among 

purification and separation processes. The body of the vessel is a horizontal container fabricated 

from schedule 40 stainless-steel pipe with a volume of about 4.4 L. The vessel is filled with 

LaNi4.25Al0.75 to approximately 70% full. The container is fully enclosed in a 4-in. schedule 10 

stainless-steel jacket (Heung 2001). 

For separating hydrogen isotopes from inert gases such as helium and nitrogen, a flow-through type 

of vessel was used. The U-shape column was built out of stainless-steel pipes up to 8 cm in 

diameter. The metal hydride for this application was palladium deposited on kieselguhr, a siliceous 

sedimentary mineral (otherwise known as diatomaceous earth). This packing material was chosen 

because of its ability to withstand hundreds of absorption and desorption cycles without decrepitation. 

The column had an inlet port and an outlet port on top of the legs of the U. Both ports were fitted with 

porous metal filters on top of the packing material to reduce the chances of clogging by fine 

particulates. After the column became saturated, it could be heated to desorb the hydrogen isotopes, 

which would then be collected as purified product (Heung 2001).

To separate the hydrogen isotopes from each other, a metal hydride chromatographic column is 

applied in a thermal cycling absorption process (TCAP) (Bradley 1983). This TCAP used the isotopic 

effect of palladium to separate the isotopes in a long column filled with palladium deposited on 

kieselguhr. The column is designed to minimize the height of the unit so it can be easily installed 

inside a glovebox. Multielement porous stainless-steel filters are fitted in the inlet/outlet ports to 

confine the packing material (Heung 2001). The feed and the outputs from the system are stored in 

metal hydride beds using LANA. Initial prototype TCAP testing using protium and deuterium was 

performed in the Advanced Hydride Laboratory at the SRTC.

Development of the TCAP began in 1980, and production-scale operation began in 1994 (Scogin and 

Poore 1994).
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3.6 TITANIUM

Tritium is periodically removed from process streams but might not be needed for several years. 

Therefore, SRS recognized a need for additional tritium storage that exceeded convenient transfers of 

gas to tanks or process hydride beds. With this need in mind, a hydride storage vessel (HSV) was 

developed for long-term (more than 5-year) storage of tritium. Several hydrides were investigated for 

use in the HSV. Titanium was chosen over depleted uranium, a LaNi4.15Al0.85 alloy, and zirconium. 

Depleted uranium has the undesirable characteristics of being pyrophoric, slightly radioactive, and 

does not retain 3

He in the metal matrix as well as other materials. Helium retention within the metal 

matrix is important because it permits storing tritium at high concentration without outgassing of 

helium leading to increased pressure on the vessel walls. Titanium hydrides are not pyrophoric in air 

and are stable in both air and water at ambient temperatures (Heung 1994a).

By weight, titanium has a greater storage capacity than the LaNi4.15Al0.85 alloy and a lower tritium 

equilibrium pressure. Titanium was chosen over zirconium because the desorption temperature for 

tritium recovery from titanium is lower than that for zirconium (Klein 2001). The basic design of the 

HSV is an upright vessel about 17 cm in diameter and 30 cm high of 6-in. schedule 40 stainless steel. 

Two pieces of stainless-steel tubing penetrate the top cap of the container body and serve as inlet 

and outlet for the flow of gases. The open ends of the tubes are welded on with porous stainlesssteel filters (2- to 5-μm pore size) to prevent titanium particles from escaping the container Heung 

2001). 

The vessel contains about 4.5 kg of titanium and can store 67 mol of tritium for 4 years before helium 

release occurs. When part of the tritium is replaced by deuterium or protium, its storage time can be 

increased proportionally. The HSV is designed to stay in storage up to the time when helium decay 

reaches the 0.3-atom ratio. If permanent storage is the objective, the assumption must be made that

all the helium will come off in the gas phase. The ultimate helium pressure limits this vessel to 

storage of 10 mol of tritium indefinitely (Heung 2001). The vessel is designed for one-time use only 

because, after heating to 600°C in the presence of tritium, the stainless-steel wall of the vessel will

most likely be saturated with tritium at the equilibrium level with the tritium pressure. This can cause 

surface contamination and weaken the vessel wall as well (Heung 1994a).

3.7 PALLADIUM

Production-scale separation of tritium from other hydrogen isotopes at SRS has been accomplished 

by several methods. These methods include thermal diffusion (1957 to 1986), fractional absorption 

(1964 to 1968), cryogenic distillation (beginning in 1967), and TCAP (beginning in 1994). In each of 

these processes, confinement systems have been used to minimize tritium release to the environment 

and to protect workers from tritium exposure. These systems typically consist of a primary 

containment system such as the material vessel, a secondary confinement such as a glovebox, and 

accessory systems for removing contaminants from the secondary containment atmosphere in the 

event of a leak from the primary containment. Depending on the process, these accessory systems 

can include stripper systems, getter systems, or some combination of the two.

The primary functions of a stripper system are to concentrate and recover the tritium, reduce the 

background level of tritium in the confinement system, and reduce worker exposure. The 

conventional and most widely used method of removing tritium from confinement atmosphere gas is 

catalytic oxidation followed by absorption in a drying agent such as zeolite. The baseline zeolite

molsieve recovery technology at SRS uses a magnesium bed to decompose water. In conventional 

oxidation-absorption stripping technology, the feed gas is processed through a heated catalyst bed 

(350°C to 500°C) to convert elemental hydrogen to oxides (water). The gas is then cooled to 20°C to 

30°C to absorb the waters in a dessicant, moisture-absorbing material, typically a synthetic zeolite. 

Tritium in the form of oxide is thus removed from the gas and trapped in the zeolite (Birchenall 1992). 
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To reclaim the tritium, the zeolite is heated, evolving the trapped water, which is then passed over an 

elevated temperature uranium bed. The uranium bed oxidizes, cracking the water released from the 

zeolite bed. The oxygen from the water is retained on the uranium bed, and the resulting hydrogen 

isotope mixture can be reprocessed (Birchenall 1992). At SRS, the catalyst bed has typically been a 

palladium-aluminum or palladium-silver alloy. However, because this material is only used as a 

high-temperature catalyst in the oxidation of elemental tritium, metal hydrides are not formed unless 

inadvertent cooling of the diffuser membrane occurs in the presence of hydrogen (Clark et al. 1995).

The earliest diffusers at SRS were single-stage, water-cooled units designed specifically for the 

Tritium Facilities by DuPont. Palladium is able to store a large atomic percentage of hydrogen at 

room temperature and allows hydrogen to diffuse with a high mobility. However, one drawback of 

palladium is that, as it is for many metals, the addition of hydrogen to a metal adversely affects the 

material’s mechanical properties by reducing ductility and increasing susceptibility to fatigue and 

corrosion, which necessitates the backing of palladium with a hydrophobic solid such as kieselguhr or 

the development of a palladium alloy. An application in which palladium that requires mechanical 

resilience is useful is as a hydrogen diffusion membrane. By filtering hydrogen gas through a 

membrane of palladium, high-purity hydrogen can be obtained because the palladium allows for high 

diffusion of hydrogen atoms while blocking the transmission of impurities (Hale et al. 2013).

After 1958 these units were redesigned to be a “multistage palladium diffuser” (MPD). Each stage of 

the MPD consisted of one pure palladium diffuser tube wound in a coil supported by a ceramic core in 

its own flanged section, and the stages were connected by stainless-steel jumpers. Each stage was 

contained by a water-cooled stainless-steel shell and cooling jacket, and the stages were connected 

using O-ring seals. Water-cooling of the shell reduced diffusion losses through the shell and 

prevented heat damage to the O-rings. The last of the MPD stages was taken out of service in 1985 

(Motyka et al. 1995).

Diffusers manufactured by Resource Services Incorporated (RSI) came into service around 1983. 

These commercially available units used a 75% palladium, 25% silver alloy for the diffusion coil 

(Motyka et al. 1995). Alloying palladium with another metal offers the opportunity to alter the 

properties of the hydride and improve its overall performance. For instance, additions of sliver greatly 

improve the mechanical resilience of palladium hydrides by suppressing the miscibility gap at room 

temperature. This gap results in the separation of the material into a low hydrogen (α) phase, and a 

high hydrogen (β) phase. The lattice mismatch between these phases results in considerable strain 

and plastic deformation within the material. Depending on the silver composition, the hydrogen 

diffusion rate can also be increased in relation to pure palladium. However, the addition of silver also 

decreases the hydrogen solubility and results in a trade-off of properties with changing compositions 

(Hale et al. 2013). 

The RSI diffuser consisted of five coiled palladium-silver permeation tubes mounted in parallel in a 

single chamber to reduce the pressure drop associated with high tube-side gas flows (Motyka et al. 

1995). 

Early in the 1990s, concerns about the cost of maintenance of palladium silver diffusers, the 

production of tritiated water by the use of catalytic stripper systems, and the need for disposal of 

uranium beds as waste prompted investigations into the use of metal hydride getter beds for tritium 

removal. Aside from the production of tritiated water, the problem with the uranium bed cracking 

system is that the uranium bed is consumed in the process and cannot be regenerated because the 

temperatures necessary to decompose the stable uranium oxide would result in a sintered uranium 

product with greatly reduced surface area for subsequent water cracking (Birchenall 1992). 

In January 1994, consolidation of operations from some of the older buildings into newer facilities

began. Part of the gas purification process in this new facility was a flow-through bed using a 

U-shaped column filled with palladium on kieselguhr. These beds allowed hydrogen to be absorbed 
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and desorbed without the heat necessary to decompose uranium, which permitted the beds to be 

reused (Clark et al. 1995).

Assuming occasional inadvertent cooling of the palladium, failure of the coils due to hydride formation

might be anticipated. No record of such a failure at SRS has been found. Heaters, heater controllers, 

and thermocouples have all failed on occasion, but apparently insufficient hydrogen was present in 

the palladium coils to cause hydriding at the lowest temperatures it reached. Multiple thermocouples 

are normally used, so that the control function of a failed thermocouple can be replaced by a backup 

(Motyka et al. 1995).

3.8 INTENTIONALLY FORMED ZIRCONIUM

A newer system, the Tritium Process Stripper, uses metal hydride getters and diffusers to remove 

tritium from process waste streams without the formation of tritium oxide. To strip tritium from a 

process stream, two getters are used in series: a sacrificial getter that reacts with oxygen and cracks 

water to convert all hydrogen isotopes to the elemental form, followed by a second getter that getters 

hydrogen as a hydride (Hsu and Oji 1997). A commercially available zirconium-manganese-iron alloy 

getter material, SAES St 909, is used in a bed as the sacrificial getter to absorb oxygen and reduce 

hydrogen-containing compounds to elemental hydrogen. The primary getter chosen for the process 

was SAES St 198 after testing indicated that the material could strip hydrogen from nitrogen or inert 

streams down to the desired part per million level (Clark et al. 1995). When the St 198 bed becomes 

saturated, it is heated to desorb the hydrogen isotopes, which then undergo isotopic separation (Hsu 

and Oji 1997).

SAES St 909 was originally developed for dissociating (cracking) tritiated water for the recovery of 

tritium. Various zirconium alloys were tested for water vapor sorption rates and hydrogen “pick-up” 

rates, and the zirconium-manganese-iron alloy was chosen as the material with the best combination 

of water-cracking rate and low hydrogen pick-up. The zirconium-manganese-iron alloy was combined 

with an aluminum binder for pellet formation and sold as SAES St 909. St 909 was shown to crack 

methane, carbon monoxide, carbon dioxide, ammonia, and oxygen in helium streams.

In 1999, SRS started testing St 909 to determine its methane-cracking performance with different 

methane concentrations, with different impurities, and in carrier gases of helium, hydrogen, and 

nitrogen. Test programs using bench-scale (6-gram), pilot-scale (500-g), and full-scale (5,300-g) 

St 909 beds were conducted (Klein and Holder 2005). 

The St 198 metal getter, having a reported composition of 76.5% Zr and 23.5% Fe, is unique from a 

number of other metal hydride-based getter materials in that it is reactive with hydrogen isotopes, 

oxygen, water, and other gases, while remaining relatively inert to nitrogen. This property allows the 

St 198 to be used in nitrogen glovebox atmospheres (Nobile et al. 1994).

3.9 ZIRCONIUM TRITIDES FROM FUELS

The most important movement of tritium in a light-water reactor (LWR) is diffusion of a portion of 

tritium from the fuel into the Zircaloy cladding, where it is trapped as a hydride. The portion of tritium 

trapped in the cladding is typically 15% to 60% of the tritium produced in the fuel (IAEA 2004). To 

recover target components from the zirconium or thorium-based hydride fuels by solvent extraction, 

the fuel and Zircaloy cladding must be dissolved and suitable feed solution prepared. Because 

zirconium resists dissolution by nitric acid, the choice of reagents to dissolve hydride fuels is limited 

(Mozin 2008). In an LWR fuel reprocessing plant using the plutonium-uranium extraction (PUREX) 

process, after chopping the fuel elements and fuel dissolution in nitric acid, about 60% of the total 

amount of tritium is retained in the cladding. The bulk of the remainder is in the form of HTO and nitric 

acid, which is distributed through the various streams in the process (IAEA 2004).
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Alloys of zirconium (i.e., Zircaloy) have been used as cladding for nuclear fuel because of their low 

thermal neutron absorption cross section and their relatively good resistance to corrosion and 

radiation damage. Zircaloy cladding makes up roughly 25% by mass of used nuclear fuel in the 

United States (Crowder et al. 2010). The alloying elements and composition are varied to provide the 

proper corrosion resistance while maintaining strength and ductility to prevent failures during reactor 

operations. For example, it was discovered that the presence of nickel in Zircaloy-2 (Zr-2) promoted 

absorption of hydrogen. Because pressurized-water reactors (PWRs) use an overpressure of 

hydrogen in the reactor coolant, Zircaloy-4 (Zr-4) was developed in which the nickel was replaced with 

additional iron to maintain corrosion resistance. Zr-2 is used exclusively for boiling-water reactors 

(BWRs) and has as its major constituents 97.963% Zr, 1.6% Sn, 0.15% Fe, 0.1% Cr, and 0.05% Ni. 

In the major BWR fuel rod types, the thickness of the cladding ranges from 0.03 to 0.037 in. Zr-4 is 

used in PWR fuel rods and has as its major constituents 97.911% Zr, 1.6% Sn, 0.225% Fe, 0.125% 

Cr, and 0.002% Ni. In the major PWR fuel rods types, the thickness of the cladding ranges from 

0.0225 to 0.03 in. (Bond, Mailen, and Michaels 1992).

Zirconium alloy cladding materials resist attack by nitric acid, which is the most widely accepted 

solvent for recovery of uranium and plutonium from fuels. To overcome this resistance, the cladding 

must be removed or altered in such a way that the fuel can be dissolved and used in a process feed 

stream. The chop-leach process has been employed in commercial reprocessing plants for the 

recovery of uranium and plutonium from spent LWR fuels. In these plants, the fuel is dissolved from 

short segments of chopped fuel rods by an aqueous nitric acid solution, then the uranium and 

plutonium are recovered by solvent extraction, using the PUREX process at first and later the 

uranium-extraction UREX process, as a solution of their nitrate salts (Bond, Mailen, and Michaels 

1992). In this process, tritium that is bound in the cladding in the form of zirconium tritide follows the 

cladding to its eventual disposal site.

Electrolytic dissolution of Zircaloy in nitric acid was recognized as a possibility with the virtue of 

allowing solvent extraction by conventional means in conventional stainless-steel equipment. A 

5,000-ampere dissolver was installed at the Semiworks in early 1961. The design was based on the 

“metal contact” concept in which the fuel assemblies to be dissolved were placed in an anodic 

niobium basket inside a niobium cathode. In late 1961, the dissolver was converted to a “liquid 

contact” system in which the fuel assemblies to be dissolved were suspended between a platinumclad niobium anode and a niobium cathode (Caracciolo and Kishbaugh 1964). 

Under electrolysis, the majority of the Zircaloy cladding is converted to an oxide that falls to the 

bottom of the dissolver as a sludge, with 15% remaining in solution. Two methods of sludge removal 

were investigated at SRS and found satisfactory. In the first method, a trough was installed below the 

electrodes to catch the sludge. The sludge and dissolver solution were continuously pumped from the

trough to a decanter or a 200-mesh screen canister where the sludge settled and the solution 

returned by gravity to the dissolver. In the second method, the sludge was allowed to collect in the 

dissolver and, at the completion of the run, was rinsed with dilute acid to remove absorbed uranium. 

The sludge was then flushed toward the jet suction by water sprays in the bottom of the tank and 

removed as a slurry. The electrolytic dissolver operated through 1979, when most of the processable 

fuel was exhausted (Bebbington 1990).

Several research groups have measured the tritium content of irradiated fuels and Zircaloy hulls by 

various methods with varying results. Measurements of 0.5- to 0.6-g Zr-2 samples made at SRS 

showed a range of 67 to 99 μCi tritium/g hull with an average of 86 μCi/g (Crowder, Laurinat, and 

Stillman 2010).

The use of Zircaloy as a cladding for fuel rods protects the fuel from corrosion by the cooling water, 

but it can degrade over time as a result of the hydrogen that permeates the crystalline structure of the 

metal. Helium is a product of the decay of tritium and, in contrast to tritium, is a very stable atom that 
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shows an extremely low solubility in metals. Due to its insolubility, accumulation of helium atoms can 

produce highly pressurized bubbles (Laesser 1989). Metal tritide films of five different elemental 

tritides were observed by Beavis and Miglionico (1972) at Sandia National Laboratories over a 5- to 

6-year period.

Once a critical ratio of helium to metal (0.25 to 0.35) is reached, the rupture of pressurized bubbles 

near dislocation networks with continuous paths to the surface results in sudden helium release. 

Abrupt increases in helium released from tritides are accompanied by flaking or blistering of the 

surface. This damage from helium buildup can cause small particles to break off the tritide surface 

(Schober, Tinkaus, and Lasser 1986; Schober 1988; Schober and Farrell 1989). Studies have shown 

a considerable detachment of tritide particles due to the formation and rupture of the bubbles. 

Detached particles varied in size from 1 to 50 μm in diameter (Beavis and Miglionico 1972).

While the production of hydrides in fuel cladding is generally considered undesirable due to 

hydride-induced fracturing, researchers have sought to capitalize on the storage capacity of tritium in 

zirconium hydrides by placing the hydride in contact with the tritium as soon as it is produced. In 

December 1998, DOE announced that commercial reactors would be the source for new tritium 

production. Assemblies of tritium-producing burnable absorber rods (TPBARs) are irradiated in a 

commercial LWR to produce tritium. TPBARs are similar in size and nuclear characteristics to 

standard, commercial LWR, stainless steel-clad burnable absorber rods. The internal components 

have been designed and selected to produce and retain tritium. Within the stainless-steel cladding is 

a metal “getter” tube that encircles a stack of annular ceramic pellets of lithium aluminate. The pellets 

are enriched with 6

Li (DOE 2016). 

When irradiated in the LWR, the 6

Li pellets absorb neutrons, which simulates the nuclear 

characteristics of a burnable absorber rod, and produce tritium. The tritium chemically reacts with the 

metal getter, which captures the tritium as a metal hydride. The irradiated TPBARs are removed and 

shipped to SRS. At SRS, the tritium is extracted from the irradiated TPBARs and purified through a 

number of chemical and physical separation steps (Hsu and Heung 1997).

In October 2003, the first TPBARs were inserted into the Tennessee Valley Authority (TVA) Watts Bar 

Unit 1 reactor for irradiation. The first shipment of irradiated TPBARs arrived at SRS in August 2005 

and was stored, awaiting completion of the processing facility. The first extraction of tritium from 

TPBARs was successfully completed in January 2007.

To meet design limitations on rod internal pressure and burnup of the lithium pellets, the amount of 

tritium production per TPBAR is limited to a maximum of 1.2 g over the full design life (less than 

495 equivalent full-power days). The potential release rate of tritium into the reactor coolant is subject 

to a design limit of less than 1,000 Ci/1,000 TPBARs per year. This is achieved by the combined 

effects of the metal getter tube surrounding the lithium aluminate pellets and an aluminide barrier 

coating on the inner surface of the cladding.

4.0 STABLE METAL TRITIDE HANDLING AREAS

STCs have been present in several SRS facilities. These materials are used in tritium processes as 

storage media and are the subject (or byproduct) of research activities and fuel processing.

4.1 F-AREA

In 1950 the U.S. Atomic Energy Commission authorized the Savannah River Project principally for the 

production of tritium for weapons use. Tritium processing operations have been performed at SRS 

since October 1955. Process equipment and designs were drawn from experience at the Hanford 
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Site, and tritium process flows were extrapolated from the initial tritium production conducted at the 

Los Alamos National Laboratory.

The process equipment was enclosed in a stainless-steel cabinet with panels (sometimes called a 

hood). The panels were framed and held glass windows for visual inspection of equipment. The 

panels were mounted on tracks to permit opening for access to equipment for maintenance or 

replacement. This early design proved to be ineffective in containing contamination within the 

cabinet. Soon after startup, contamination problems were evident within the process room, 

sometimes spreading to other adjacent rooms (Blackburn 1994). 

This area processed its final charge in October 1958 (DuPont 1973), after which all of the process 

lines and equipment were flushed with inert gas (Blackburn 1994). A characterization of the building 

in 1994 indicated that tritium was extracted in the facility from LiAl slugs. Mercury is known to have 

been used in the pumps, and the process piping was assumed to be contaminated with mercury 

(Scallon and England 1994). However, hydrogen is fairly insoluble in mercury, forming a hydride or 

dihydride only at extremely low temperatures, which makes the presence of mercury hydride at SRS 

extremely unlikely. The facility had made use of palladium-silver diffusers and zeolite beds, but there 

are no indications that metal tritides were fabricated or used in this area (Scallon and England 1994).

4.2 H-AREA

In H-Area, tritium processes started in 1957 (Buley et al. 1995) and continued until approximately 

2002. Processing in this area included vacuum heating of reactor targets for the extraction of tritium, 

purification by palladium diffusion, and separation by thermal diffusion and cryogenic separation 

(Rabun 2000).

The first HSV was loaded on April 3, 1996, delivered to H-Area, and installed in a monitoring manifold. 

Temperature and pressure data have been recorded from this apparatus since late May 1996. Nine 

samples were originally prepared with varying tritium and deuterium loadings for use in a long-term 

storage program. One sample has been used to collect pure tritium isotherm data on an annual basis 

(Shanahan 1998).

After 2004, personnel from SRTC performed experimental work to develop tritium processing and 

handling techniques. Tritium was pumped, compressed, purified, dissolved in, and reacted with 

various materials. A variety of STCs were deliberately made such as uranium tritide, palladium tritide, 

titanium tritide, and LANA tritide. Tritiated vacuum pump oils are known to have been formed during 

processing.

All metal hydrides used in tritium production and in research and development (R&D) are contained in 

stainless-steel vessels, which are referred to as metal hydride beds, equipped with sintered metal frit 

filters. These beds usually consist of horizontal cylindrical vessels, with the metal hydride powder 

filling about 70% of the volume. The powder occupies the lower portion of the horizontal cylinder, 

allowing some free volume above the powder for normal expansion and contraction of the powder. 

There are some applications that involve flow through the metal hydride powder. In these 

applications, the metal hydride is palladium supported on kieselguhr, the cylindrical cavity is fully 

packed, and the bed configuration is a coiled tube or a U-tube (Nobile and Motyka 1991). All tritium 

operations were performed in air-ventilated hoods. Kanne tritium air monitors were used to indicate 

airborne tritium activity in hoods and rooms. Work areas and adjacent areas that had the potential for 

contamination were monitored by smearing with filter paper and counting with liquid scintillation 

counting on a regular basis and immediately after completion of any job.

All tritium process equipment in H-Area is secondarily confined in gloveboxes or jacketed piping, the 

first large-scale tritium handling facility in the DOE complex to do so. These operations included 
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hydrogen isotopic separation, loading and unloading of tritium, shipping and storage recovery from 

zeolite beds, and stripping of nitrogen flush gas to remove tritium before stack discharge (Hsu and 

Oji 1997).

Additional operations conducted in this area include purification of tritium by palladium diffusion, 

storage of gaseous tritium, and small-scale storage of tritium using storage beds. 

A 3

He purification process in H-Area involves slowly passing the gas through a cold gold trap, then

through an ambient zeolite bed to remove any moisture from the gas stream, and finally through two 

liquid nitrogen-cooled zeolite beds connected in series to condense out the inert gases and nitrogen. 

Purified helium-3 is stored in a feed tank that is loaded via a compressor into a gas cylinder. The cold 

zeolite beds are regenerated after each transfer by allowing them to slowly warm to ambient 

temperature, allowing the inert gases and nitrogen trapped in them to be released to the waste tank

(Hsu and Oji 1997). 

In December 1998, DOE announced that commercial reactors would be the primary source for tritium 

production. In May 1999, DOE selected TVA’s Watts Bar Unit 1, Sequoyah Unit 1, and Sequoyah 

Unit 2 commercial LWRs for irradiating the DOE-supplied TPBARs. In October 2003, the first 

TPBARs were inserted in TVA’s Watts Bar reactor for irradiation. The SRS was tasked by the DOE to 

design and construct a Tritium Extraction Facility to process irradiated tritium-producing burnable 

absorber rods (TPBARS) from a commercial LWR. The intent is for the rods to provide tritium to 

support DOE requirements. DOE started funding the conceptual design for the project in 1996. The 

TPBAR was designed by Pacific Northwest National Laboratory. The TPBAR assemblies are 

irradiated in a commercial utility LWR and transported to SRS for tritium extraction and processing

(DOE 2016).

A staff of approximately 600 workers was employed during peak construction phase of this project, 

and the facility has an operations staff of approximately 100 permanent employees. Within the project 

area, the TPBARs are unloaded and the tritium gas extracted. It has a truck receiving area, cask 

decontamination area, TPBAR and waste preparation area, furnaces, and hot maintenance area, 

along with the associated gloveboxes for extraction pumps and tanks. It also includes an overhead 

crane and remote handling equipment. Another facility provides preliminary purification of the 

extracted gases.

This project celebrated completion of the nonradioactive startup testing program on February 28, 

2006. Over 700 different systems and components were successfully tested. In November 2006, the 

project successfully began radioactive operations. In February 2007, SRS completed the startup of 

the project and made the first transfer of new tritium gas to the nation’s tritium inventory.

4.3 A-AREA

Current tritium handling practices in this area are limited to relatively small quantities of tritium. With

one exception, no SMTs are intentionally formed or handled. Quantities inadvertently formed or 

incidental to handling are expected to be very small and might be mixed with other alpha and 

beta/gamma emitters, which are a much more severe radiological hazard. Operations in this area fall 

into three categories: (1) chemical analyses of materials that contain tritium, (2) mechanical testing of 

metals that contain dissolved elemental tritium, and (3) metallographic preparation and examination of 

materials that contain tritium. Of these, only the metallographic preparation appears likely to present 

any credible hazard from respirable SMTs (Howard 2000).

Several laboratories within the buildings are used or have been used for tritium work in the past. Most 

of the metal hydride development work was done “cold” with protium and deuterium (Howard 2000). 

In about 1954, bench-scale experimental work was done to investigate the use of electrolysis to 
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remove tritium from heavy water. It is conceivable that tritiated rust could have been formed on steel 

exposed to heavy-water moisture. However, tritium was present only at tracer levels in the heavy 

water used for the experiments. The project was terminated because of practical difficulties, and all 

equipment was removed shortly thereafter.

In the mid-1970s, experimental work was done in gloveboxes and hoods to develop processes for 

handling tritiated materials. The glovebox ventilation systems used heated uranium metal and copper 

metal stripper beds to remove tritium and oxygen from the containment atmosphere. It is conceivable

that metal tritides could have been formed in these beds. Mercury diffusion pumps, backed by 

mechanical roughing pumps, were employed in the processing equipment. 

All equipment, including the gloveboxes, hoods, pumps, and other process equipment was removed 

sometime in the early 1980s and sent to the burial ground. Monitoring of this area after removal of 

this equipment has revealed no indication of residual tritium contamination.

In 1987, an experimental program was initiated to study the effects of tritium exposure on the 

absorption-desorption characteristics of metal hydrides. In this program, samples of LaNi4.15Al0.85, 

LaNi4.25Al0.75 (LANA), LaNi4.15Al0.3, and palladium were loaded with tritium and placed in dormant 

storage. At periodic intervals samples were taken from storage, and the tritium absorption-desorption 

characteristics were measured. The experimental program has produced information on these

characteristics for the materials as a function of tritium exposure (Nobile and Motyka 1991).

A “cold” process demonstration facility began operation in November 1987. Materials that were used 

in the cold process included protium and deuterium but not tritium (Posey et al. 1987). The purpose 

of the cold process area was to demonstrate the metal hydride technology by integrating the various 

unit operations into an overall process (Motyka 1989). The cold process area was automatically 

controlled, with the equipment designed to be operated as an integrated process to simulate 

production, and as a series of one or more unit processes to prove out equipment. The continuous 

presence of an operator was not required during routine operations (Posey et al. 1987).

A variety of analytical measurements have been conducted on samples that have contained tritium. 

These samples have been, variously, liquids such as heavy-water reactor moderator, reactor cooling 

water, high-level waste tank supernate, or solids contaminated with the aforementioned liquids. While 

it is conceivable that tritiated solids could have been present in these samples, or that tritiated rust 

could have been formed from exposure of steel to the tritiated water, the tritium constitutes only a 

minor fraction of the radionuclides present. These other radionuclides represent a far more severe

hazard from penetrating radiation, such as 137Cs, or from inhalation or ingestion, such as 239Pu.

When the overall activity levels have exceeded the limits for handling in the other laboratories, tritiumbearing materials were handled in shielded cells. These materials have included pieces of equipment 

contaminated with reactor moderator or coolant such as tubes or fittings that have been processed to 

decontaminate them to the point where they can be handled in other laboratories. Such 

decontamination has been accomplished by washing, rinsing, or degassing. Other materials that 

have been processed in these cells include high-level waste tank sludge and supernate. While it is 

conceivable that tritiated solids could have been present in these samples, or that tritiated rust could 

have been formed from exposure of steel to the tritiated water, the tritium constitutes only a minor 

fraction of the radionuclides present in these cells, either from these materials or from contamination 

with nontritiated materials processed in other jobs.

Mechanical testing (tensile strength, fracture toughness, yield strength, etc.) of metal samples that 

have been exposed to tritium is currently conducted. The samples are bulk samples (dimensions on 

the order of inches), and no particulates are generated or handled in this laboratory. Tritium is 
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present in the samples in the form of dissolved elemental tritium. Monitoring capabilities are present 

and detailed procedures are used to conduct all tests.

Metallographic techniques are currently used to prepare samples for, and to conduct, both scanning 

and transmission electron microscopy examinations. The metallographic preparation techniques 

involve cutting, sectioning, mounting, and polishing of the samples. It is conceivable that the cutting 

or sectioning could generate particulates with dissolved tritium. The polishing preparation is 

performed wet (with water) and does not generate any airborne particulates. The water is disposed of 

to a low-level drain. Mounting embeds the samples in a liquid plastic that hardens and holds the 

sample and the process generates no particulates.

Other current analyses in this area include the preparation of air samples from SRS facilities for 

analysis of tritium and 85Kr. The samples are received in a stainless-steel cylinder that contains a 

molsieve to trap the radionuclides. The sieve is baked to capture the moisture that could contain the 

tritium. The captured moisture is analyzed by LSC, after which it is disposed of to the low-activity 

waste tank. Only very low quantities of tritium are present. Presence or formation of significant 

quantities of tritiated solids is not considered credible.

5.0 POTENTIAL FOR METAL HYDRIDE EXPOSURE

In general, readily dispersible or respirable-sized particles of metal tritides were not handled at SRS. 

With the exception of the metal hydrides that were examined within the SRTC, these materials were 

not handled outside their process containment vessels (Rabun 2000). Process beds containing 

tritium-exposed metal hydride powder were disposed of without removal of the powder from the bed. 

On rare occasion, however, it was necessary to remove some metal hydride powder from the process 

vessel in which it was originally contained (Nobile and Motyka 1991). For instance, a number of metal 

hydrogen absorbers were placed into tritium exposure to study the effect of tritium decay on the 

thermodynamics of the metal hydride system. Removal of the material from their respective process 

vessels was to be completed in a chemical hood and only after repeated cycles of evacuation, 

heating, and backfilling of the vessel with air (Nichols 1996). When these samples reached the end of 

life, they were discontinued in the program but retained for additional study. To make the samples 

available for further study, the metal powders must be removed from the sample vessels after 

procedures to remove as much tritium as possible and to passivate the metal surface with air 

(Walters 2006).

5.1 ENGINEERING CONTROLS

Because tritium is classified as a byproduct material arising from the production of a special nuclear

material, periodic inventories must be undertaken and the material must be accounted for as a waste 

product from any nuclear fuel processing operation (Baumann and MacMurdo 1977). Therefore, 

analytical determination of tritium in process streams is an integral part of the developmental program 

for alternate fuel cycle technology at SRS. All tritium inventories in process and product tanks, 

hydride and other beds, and process piping must be accounted for, which reduces the potential for an 

undocumented exposure.

As fine metal particles, metal hydrides can be very reactive to oxygen in air. Excessive air exposure 

for most metal hydrides leads to rapid oxidation that can reach red-hot temperatures. Based on this, it 

is reasonable to surmise that efforts would have been taken to limit air exposure to avoid reduction of 

the reactivity and the capacity for hydrogen absorption, as well as mitigating the potential for fire 

hazards (Heung 2001).

According to DOE, the primary condition that must be met for the potential exposure to STCs to exist 

is that tritium contamination is present at levels greater than one-tenth of the value in Appendix D of 
Document No. ORAUT-RPRT-0072 Revision No. 00 Effective Date: 01/09/2017 Page 22 of 31

10 CFR Part 835 (1,000 dpm/100 cm2). DOE guidance states that at levels less than one-tenth of this

value, the potential for significant exposures is minimal and does not warrant action beyond standard 

practices (DOE 2015). Typically, the practices that are used to prevent release of tritium gas to the 

processing areas and to the environment have also served to prevent release of tritiated particulates 

and tritiated organics. Similarly, protection of the worker from contact with tritium gas and tritiated 

water also provides a high degree of protection from tritiated particulates (Rabun 2000).

For example, when handling those materials containing tritium as the only significant radionuclide, 

work was done in air-ventilated hoods dedicated to the handling of radioactive materials. Alternatively, 

work involving limited quantities could be conducted within a radiobench, an enclosure with the same 

shape as a glove box except that in there is an open area in place of the glove panel. For sufficiently 

low levels of tritium, work could be performed on an open benchtop. If the materials contain other 

radionuclides in quantities sufficient to present a hazard from penetrating radiation, then the work was 

done in the High Level Cells. These are highly contaminated with other beta/gamma materials. The 

ventilation systems for these hoods, radiobenches, and cells might also be contaminated 

(Howard 2000).

All metal hydrides SRS has used in tritium production and R&D have been contained in stainless-steel

vessels. These process beds would have been disposed of without removal of the metal-exposed 

powders. Rather, the bed would be largely purged of residual tritium by repeatedly heating and 

evacuating the chamber and by isotopic exchange with deuterium (Nobile and Motyka 1991). In 

addition, all tritium operations in H-Area took place in one of 30 nitrogen-filled gloveboxes (Heung 

1994b). 

To reduce tritium loss to the environment, the production area uses a multilayered confinement 

system (Heung et al. 1991):

• Gloveboxes provide secondary containment. All tritium processing equipment is installed in

sealed gloveboxes; tritium transfer lines outside the gloveboxes are contained in stainlesssteel jackets; both the gloveboxes and the transfer line jackets are filled with a recirculating

nitrogen atmosphere (Horen et al. 1993).

• Primary and secondary strippers clean up the glovebox atmosphere. The nitrogen is

continuously circulated through a primary stripper system during normal operation to remove

any tritium that might have escaped the process equipment; in the event of an accidental

tritium release the nitrogen in that specific glovebox is circulated through a secondary stripper

system to speed up the cleanup.

• Purge stripper further cleans up the purge gas. Nitrogen purge from the gloveboxes is further

stripped of tritium in a purge stripper before it is released to the environment.

• Zeolite bed recovery system recovers the tritium. All the tritiated water accumulated in the

stripper system is recovered through a zeolite bed regeneration system.

5.2 SHORTFALLS OF EXISTING CONTROLS

In May 2007, there were 30 containers of legacy tritium exposure samples of several metal tritides, 

limited to palladium, LANA, titanium, and a single gram-quantity sample of a zirconium alloy hydride. 

A small sample was removed from each for further analysis and the remainder was disposed of. This 

work was performed in a glovebox, but due to the particulate form of these materials it was 

recognized that the routine job coverage practices for gaseous tritium might not be adequate. For 

example, portable tritium air monitoring instruments are not capable of detecting the particulate tritides 
Document No. ORAUT-RPRT-0072 Revision No. 00 Effective Date: 01/09/2017 Page 23 of 31

 

because the filter should remove almost all of them before they reach the chamber (Moxley 2007). 

This shortfall can be largely overcome by counting the filters using LSC.

There are additional shortfalls for monitoring air samples using LSC, however, that are not specific to 

SRS. True activity intakes can be uncertain due to self-absorption of the weakly penetrating beta 

particles by the hydride material itself. In addition, activity collected and measured on filters tends to 

overrepresent the intake because of capture of nonrespirable particulates and dissolution in 

scintillation cocktail. The final shortfall to consider is self-absorption of tritium beta radiation due to 

dust loading of the filter (BWXT 2000).

6.0 DOSIMETRY CONSIDERATIONS

Particulate SMTs can be ingested or inhaled and result in radiation dose. After ingestion, SMTs lose 

some fraction of their tritium as HTO in stomach digestive fluids. Some SMTs lose essentially all of 

the tritium and others lose very little, depending on the solubility of the specific material in the GI tract. 

In any case, the only significant dose after ingestion of SMTs is due to the released HTO. Because 

HTO is rapidly assimilated physiologically, this dose component is readily assessed via urine 

bioassay. 

After inhalation, however, metal tritides can be deposited in the lung. Once deposited, the tritium 

bound to the particulate is removed from the body via two mechanisms:

1. Dissolution of tritium from the particulate and absorption into the body as HTO and removal 

through urine, expired air, and perspiration.

2. Mechanical transport of the particulate itself to the GI tract and removal through feces. Inhaled 

SMTs removed to the gut cause dose in the same manner as ingested SMTs as described 

above. Removal of these materials from the lung through mechanical transport can be slow 

depending on the solubility of the individual tritide. SMTs can therefore reside in the lung for a 

considerable amount of time and result in lung dose.

6.1 SOLUBILITY

The dissolution rate of particles that are deposited in the respiratory tract is a major factor governing 

retention and translocation of their constituents to other organs in the body. Radioactive particles that 

dissolve slowly in lung fluid tend to remain in the lung tissue for long periods, and the biological effects 

of the particles are usually confined to the region near the deposition site (i.e., the lung tissue). On 

the other hand, highly soluble materials, such as HTO, that deposit in the respiratory tract are 

translocated to other organs and body fluid more readily and might therefore have a shorter retention 

time in the body.

The current understanding of some metal tritides is incomplete because the rate at which the material 

dissolves in the human respiratory tract is not known. DOE guidance is applied for the few metal 

tritides that have been studied. However, all the other metal tritides are currently grouped together 

and assumed to have “slow” dissolution behavior. This assumption might lead to overly conservative 

estimates of the dose an individual would receive in the event of an intake, which makes it desirable 

to characterize the dissolution rate of metal tritides of interest (Farfan et al. 2011).

Dissolution rates of some metal tritides have been measured by immersing the particles in various 

aqueous solutions and by determining the amount of dissolved radionuclides in the solution. For 

example, a sample of tritiated LaNi4.25Al0.75 similar to that used at the SRS Tritium Facilities was 

analyzed to estimate the particle size distribution of this metal tritide powder and the rate at which it 

dissolves in the human respiratory tract after inhalation. Samples were placed in serum ultrafiltrate 
Document No. ORAUT-RPRT-0072 Revision No. 00 Effective Date: 01/09/2017 Page 24 of 31

 

(SUF), an aqueous solution that contains various salts, and were enclosed in a static dissolution 

system to determine the dissolution rates. The in vitro dissolution apparatus measured the tritium in 

HT and in solution to account for the fact that the dissolved tritium in metal tritide could exist in HT or 

could exchange with protium in the water to form HTO.

The results of this experiment indicate that more than 99% of the tritium was released in 10 days as a 

result of the LANA matrix dissolving in the SUF and the tritium diffusing or offgasing from the matrix. 

This behavior is consistent with a dissolution type F (Farfan et al. 2011).

In a similar experiment, titanium tritide samples were ground into powders that were examined in an 

optical microscope, and particle size was determined using an image analyzer. SUF was used to 

determine the dissolution rates of metal tritides within a static dissolution system. The in vitro 

dissolution apparatus was designed to measure the tritium in the gas phase and in solution. 

Retention of tritium in the lung in animal studies followed a bi-phase exponential decay with half-lives 

of 0.81 and 66 d. In rats, 37%, 29% and 16% of the tritium were excreted by urine, feces, and expired 

air after 121 d, respectively. Tritium associated with titanium tritide was eliminated in this study with an 

average biological half-life of 50 d, suggesting that titanium tritide may be classified with an ICRP 

absorption type of M (Cheng et al. 1999). Subsequent research confirmed that the dissolution rate of 

metal tritides can be expressed as a two-component decay curve representing the fast and slow 

phases of retention. The updated fast phase for titanium tritide was found to yield a 1-day biological 

half-time, with the slow phase yielding a 33-day half time. These values correspond to an ICRP 

absorption type of somewhere between F and M (Zhou 2004). 

There is some disagreement over the solubility of zirconium hydride. Using the same experimental 

procedure, zirconium powder was purchased and ground to respirable sizes (1 to 3 μm) in a ball mill 

at Lovelace Respiratory Research Institute. A sample of the material was then tritiated at the Pinellas 

Plant, after which the fine powders were examined and electron photomicrographs were taken under 

a transmission electron microscope to determine the activity median aerodynamic diameter of the 

particles. Similar to the procedures in the LANA solubility studies, samples of the hydride were placed 

in SUF and enclosed in a static dissolution system to determine the dissolution rates. The in vitro 

dissolution apparatus measured the tritium in the gas phase (HT) and in solution (Zou et al. 2010).

The dissolved pattern was characterized by an initial fast-rate release of tritium that slowed with time. 

Total dissolved tritium in 71 days accounted for 37.2% ±1.3% of the initial tritium present in the 

particles; 65% of the total dissolved tritium was released after the first month of dissolution. The 

retention half-times for tritium as calculated from the dissolution rate constants are 42 and 385 days

for the fast and slow phases of the retention function, with an overall half-time of 114 days. The 

results indicate that the ZrT is a poorly soluble material (Zou et al. 2010). A similar study by Zou and 

Cheng (2004), however, concluded that the dissolution rate of ZrT ranked between Types M 

(moderate) and F (fast) materials.

For materials that have not explicitly been tested for solubility, the tendency for tritium to desorb from 

the material can be estimated with knowledge of the enthalpy of the material. Very stable hydrides 

have a large negative reaction enthalpy and dissociate only at high temperatures. A summary of the 

absorption types of the metal tritides considered in this paper is presented in Table 6-1. Note that 

none of the materials considered have an ICRP absorption type of S, indicating a very insoluble 

material. 
Document No. ORAUT-RPRT-0072 Revision No. 00 Effective Date: 01/09/2017 Page 25 of 31

Table 6-1. Absorption types of 

representative metal tritides.

SMT

Absorption 

type

LANA F

Palladium F

Palladium-aluminum M

Palladium-silver F

Titanium M

Uranium F

Zirconium M

Zirconium-iron-chromium M

6.2 SAMPLE SELF-ABSORPTION

The radiation dose to tissue from metal tritide particles depends on their solubility and retention in the 

body. In each tritide particle, a portion of the beta particles from decay of tritium is absorbed by the 

metal matrix and therefore cannot contribute to absorbed radiation dose to tissue. The betas 

originating inside a particle can be absorbed before reaching the particle surface, a phenomenon 

referred to as self-absorption (Lantz and Steward 1988). To determine dose to the lung from metal 

tritide particles imbedded in the lung, and to correct LSC of bioassay samples that contain metal tritide

particles, the self-absorption of the tritium betas by the tritide particle should be considered. Because 

the tritium betas are of such low energy (average 5.7 keV), self-absorption in a metal tritide particle 

even as small as 1 μm is significant. The self-absorption factor (SAF), the fraction of beta particles 

that escapes from a particle, can be found for spherical particles by numerical integration 

(Tsoulfanidis 1991, as referenced in Kropf et al. 1998). Similarly, when analyzing air monitoring 

samples for metal tritide using LSC, it is necessary to consider SAF because STCs do not dissolve 

appreciably and release tritium to LSC cocktail. Only beta radiation that escapes the particles in LSC 

cocktail is available for detection. The SAF is used to estimate the “true” STC activity that would be 

“observed” when samples are counted via LSC.

Research has suggested that bremsstrahlung photons, created when beta particles are self-absorbed, 

are not expected to be a significant contribution to dose from tritiated particulates (DOE 2004).

6.3 DETERMINATION OF INTAKES FROM EXISTING DATA

To determine an individual’s dose from tritium, an intake assessment usually relies on bioassay. 

However, due to the insoluble nature of SMTs and the fact that an established fecal bioassay protocol 

is not currently available for these materials, the intake amount can better be determined from 

measurements of SMT concentration in the air or from assumed resuspension of surface 

contamination.

6.3.1 Air Sampling Data

Air monitoring can be used to estimate intake directly, as opposed to the indirect bioassay methods. 

Particle size and dissolution rate variabilities have minor effects on intake determination via air 

monitoring. Intake is considered to be proportional to the true activity captured on the filter of an air 

sampling device. Particulates tend to shield their tritium beta activity by self-absorption of the beta 

radiation within the mass of particulate. Observed activity on a filter sample, measured by suspending

particulates from the filter into LSC solution, therefore underrepresents the true activity available for 

deposition to the lung (i.e., intake). However, when tritiated particulate intake is defined in terms of 

observed activity (and when the dose conversion factor is correspondingly defined in terms of 

observed activity intake), the uncertainty in that observed intake essentially disappears because selfabsorption is no longer a concern.
Document No. ORAUT-RPRT-0072 Revision No. 00 Effective Date: 01/09/2017 Page 26 of 31

 

6.3.2 Surface Contamination Data

Removable surface contamination can produce elevated airborne contamination due to resuspension. 

The fraction of surface contamination that can be resuspended is highly dependent on the type of 

activities in the contaminated area. The most widely used method for quantifying tritium 

contamination on smear and air samples is LSC. Although this technique gives acceptable results for 

tritium forms that are soluble in LSC cocktail (e.g., HTO), it generally underestimates the true tritium 

activity if STCs are present. Therefore, consideration of sample self-absorption is necessary.

7.0 CONCLUSION

Various metal tritides are encountered at SRS Tritium Facilities. In the early 1980s, Savannah River 

Laboratory began a major program to develop and use metal hydrides in tritium production. Metal 

tritides are metals that absorb and store tritium atoms in the crystalline structure of the metal. When a 

metal tritide is inhaled, the lungs retain the material, which slowly releases tritium. This release of 

tritium occurs as the metal tritide particles dissolve and the tritium diffuses out. Once the tritium is 

released from the metal tritide particle, it is assumed to be converted to HTO and subsequently

behave according to the HTO biokinetic model.

Metal hydrides turned out to be ideally suited for tritium handling and processing. These metals 

include uranium, palladium, titanium, lanthanum-aluminum-nickel tritide, mischmetal, and zirconium. 

In general, tritides at SRS can generally be characterized as readily soluble metal tritides. 

Evaluations of the physical and chemical properties of these tritium compounds have shown that 

uranium and palladium do not bind tritium in a manner significantly different from tritium oxide. 

Therefore, they are not considered STCs. The remaining metals form compounds that cannot be 

adequately assessed using the tritium oxide urine bioassay model and therefore are classified as 

STCs.

As stated in OTIB-0066, in the vast majority of occupational exposures to tritium it is not possible to 

identify the tritium compound taken into the body based on the observed urinary excretion. Therefore, 

the selection of the appropriate tritium compound in an intake evaluation must usually be based on 

process knowledge of the source terms in the workplace. The material presented in this paper, and 

the classified notes associated with this paper, describes the types of SMTs used at SRS, the 

locations at which these materials could be encountered, and the time frames during which these 

materials were handled. This information, when coupled with existing tritium bioassay data and the 

methods presented in OTIB-0066 should be sufficient to bound metal hydride doses at SRS.
Document No. ORAUT-RPRT-0072 Revision No. 00 Effective Date: 01/09/2017 Page 27 of 31

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