uHH h

Waflonal Bureau of Standards

Library, E-Ol Admin. Bldg.

SEP 4 1970 NBS TECHNICAL NOTE 508

Activation Analysis

%/wwll wl urn Summary of Activities

July 1968 to June 1969
NATIONAL BUREAU OF STANDARDS

I he National Bureau of Standards ' was established by an act of Congress March 3, 1901. Today,

in addition to serving as the Nation's central measurement laboratory, the Bureau is a principal

focal point in the Federal Government for assuring maximum application of the physical and

engineering sciences to the advancement of technology in industry and commerce. To this end

the Bureau conducts research and provides central national services in four broad program

areas. These are: (1) basic measurements and standards, (2) materials measurements and

standards, (3) technological measurements and standards, and (4) transfer of technology.

The Bureau comprises the Institute for Basic Standards, the Institute for Materials Research, the

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THE INSTITUTE FOR BASIC STANDARDS provides the central basis within the United

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divisions:Applied Mathematics—Electricity—Metrology—Mechanics—Heat—Atomic and Molecular Physics—Radio Physics -—Radio Engineering - —Time and Frequency - —Astrophysics -—Cryogenics.

-

THE INSTITUTE FOR MATERIALS RESEARCH conducts materials research leading to im- proved methods of measurement standards, and data on the properties of well-characterized

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THE OFFICE FOR INFORMATION PROGRAMS promotes optimum dissemination and

accessibility of scientific information generated within NBS and other agencies of the Federal Government; promotes the development of the National Standard Reference Data System and a system of information analysis centers dealing with the broader aspects of the National Measure- ment System, and provides appropriate services to ensure that the NBS staff has optimum accessibility to the scientific information of the world. The Office consists of the following

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UNITED STATES DEPARTMENT OF COMMERCE

Maurice H. Stans, Secretary

NATIONAL BUREAU OF STANDARDS • Lewis M. Branscomb, Director

NBS TECHNICAL NOTE 508

ISSUED JULY 1970

Nat. Bur. Stand. (U.S.), Tech. Note 508, 147 pages (July 1970)

CODEN: NBTNA

Activation Analysis

Section:

Summary of Activities

July 1968 to June 1969

Edited by Philip D. LaFleur

Activation Analysis Section

Analytical Chemistry Division

Institute for Materials Research

National Bureau of Standards

Washington, D.C. 20234

NBS Technical Notes are designed to supplement the Bureau's regular publications program. They provide a means for making available scientific data that are of transient or limited interest. Technical Notes may be

listed or referred to in the open literature.

For sale by the Superintendent of Documents, U.S. Government Printing Office, Washington, D.C, 20402

(Order by SD Catalog No. CI 3.46:508), Price $1.25 .

FOREWORD

The Analytical Chemistry Division was established as a separate division at the National Bureau of Standards on

September 1, 1963, and became part of the Institute for

Materials Research in the February 1, 1964, reorganization.

It consists at present of nine sections and about 100 technical personnel encompassing some 57 different analytical

competences from activation analysis and atomic absorption to

vacuum fusion and x-ray spectroscopy. These competences, and

in turn the sections which they comprise, are charged with

research at the forefront of analysis as well as awareness of

the practical sample, be it standard reference material or

service analysis. In addition it is their responsibility to

inform others of their efforts.

Formal publication in scientific periodicals is a highly

important output of our laboratories. In addition, however,

it has been our experience that informal, annual summaries of

progress describing efforts of the past year can be very valuable in disseminating information about our programs. A word

is perhaps in order about the philosophy of these yearly

progress reports. In any research program a large amount of

information is obtained and techniques developed which never

find their way into the literature. This includes the

"negative results" which are so disappointing and unspectacular but which can often save others considerable work. Of

importance also are the numerous small items which are often

explored in a few days and which are not important enough to

warrant publication—yet can be of great interest and use to

specialists in a given area. Finally there are the experimental techniques and procedures, the designs and modifications

of equipment, etc., which often require months to perfect and

yet al^ too often must be covered in only a line or two of a journal article.

in
Thus our progress reports endeavor to present this information which we have struggled to obtain and which we feel

might be of some help to others. Certain areas which it

appears will not be treated fully in regular publications are

considered in some detail here. Other results which are

being written up for publication in the journal literature

are covered in a much more abbreviated form.

At the National Bureau of Standards publications such as

these fit logically into the category of a Technical Note.

We plan to issue these summaries for all of our sections.

The following is the third annual report o'n progress of the

Activation Analysis Section.

W. Wayne Meinke, Chief

Analytical Chemistry Division

IV
PREFACE

The Activation Analysis Section has the responsibility

to develop and apply measurement techniques for the quantitative and qualitative analysis of materials through the

use of nuclear reactions.

The basic activities of the Section may be divided into

two general areas, the research and development area and

the sample area. The Section has a very real responsibility

for the development of new techniques, and the refinement of

existing techniques, to increase the applicability of activation analysis and to reduce the cost of analysis to become

more competitive with non-nuclear methods of analysis.

A very important part of the Section's activities is

involved in the certification of Standard Reference Materials

in cooperation with the other Sections in the Analytical

Chemistry Division. The Office of Standard Reference Materials of the National Bureau of Standards is currently developing, or has issued, Standard Reference Materials of very high

purity which are well characterized for their trace element

content. The Activation Analysis Section has made a significant contribution in the certification of these standard

materials

Activation analysis is particularly suited for the analysis of trace and ultra-trace components of materials because

of the very high sensitivity for many elements, the lack of a

reagent "blank" in most analyses, and the opportunity for

excellent accuracy when carefully done.

The Activation Analysis Section is composed of six groups

three of which are oriented to a specific source of nuclear

radiation: the reactor, LINAC , and Cockcroft-Walton generator groups, a group for the development of radiochemical separations, and, in recognition of the increasing emphasis on the

analysis of biological materials in the Section, we are activating a project oriented to this important sample area. The

v
Sixth project , Instrumentation Techniques and Automation has,

at present, no personnel assigned. Considerable interaction

among the groups is encouraged so that activities of common

interest are worked upon jointly in the several groups. In

addition, the Radiochemical Analysis Section retains groups

in nuclear instrumentation and nuclear chemistry which consult

both Sections.

A roster of the groups in the Section is listed in Part 7.

In the past the Section has enjoyed the presence of scientists

from abroad, and from within the United States, who have

worked for one to two years in the Section, and have contributed to a very stimulating environment. It is hoped that

we may continue to utilize this program in the future.

In order to specify adequately procedures, it has been

necessary occasionally to identify commercial materials and

equipment in this report. In no case does such identification

imply recommendation or endorsement by the National Bureau of

Standards, nor does it imply that the material or equipment

identified is necessarily the best available for the purpose.

Philip D. LaPleur, Chief

Activation Analysis Section

Analytical Chemistry Division

VI
TABLE OP CONTENTS

1. INTRODUCTION

2. LIAISON AND GENERAL FACILITIES IN THE NBSR. . .

A. Introduction

B. Liaison

C. Pneumatic Tube System

1. Instrumentation and Operation

2. Safety Systems

D. Manual Irradiation Facility

E. Status of Existing 1 1/2 in. Diameter

Horizontal Pneumatic Tube Systems

F. Progress of Irradiation Facilities in NBSR.

G. Measurement of Temperature and Neutron Flux

Characteristics of the NBSR Pneumatic Tube

Irradiation Facilities RT-2 and RT-3. . . .

1. Design Information on Pneumatic Tubes

RT-2 and RT-3

2. Experimental

3. Results

3. ACTIVATION ANALYSIS WITH A NUCLEAR REACTOR. . .

A. Introduction

B. Facilities

1. Modifications to a Beta-Gamma-Gamma

Sum Coincidence Spectrometer

2. Counting" Room

C. Research Activities

1. Activation Analysis for Molybdenum in

the Presence of Large Amounts of

Tungsten

2. Ultrapure Reagents: Use of the

Freezing Technique in the Quest for the

Absolute Blank

vii

Page

2

4

4

4

5

7

10

11

13

14

14

15

15

16

22

22

22

22

25

26

26

29
TABLE OF CONTENTS (cont.)

Page

3. Analysis of Ancient Stained Glasses

by Neutron Activation Analysis 34

4. Rapid Radiochemical Separation of

Uranium 35

D. Analysis of Standard Reference Materials

(SRM) 37

1. Activation Analysis of SRM Steels Using

Group Radiochemical Separations .... 37

2. The Determination of Manganese and

Vanadium in SRM Nos . l6b , 115? 1265 x 1261,. 36b, 20g ^0

3. The Analysis for Lanthanum, Tantalum

and Selenium in SRM Special Iron No.

1265 43

4. The Determination of Homogeneity of

Picogram Quantities of Sodium by

Neutron Activation Analysis

5. The Determination of Impurities in

Clinical Standards 49

6. The Determination of Gold in the SRM

Cobalt-Aluminum Flux Wires 52

7. Analysis of Simulated Lunar Glasses . . 55

8. Determination of Tungsten in SRM 480- •

E. Service Analyses 60

44

1. Determination of Chlorine in AI2O3 and

TaaOs by Activation Analysis

2. Determination of Selenium in Electrodeposited Copper

3. The Determination of Selenium by

Neutron Activation Analysis 63

60

60

4. ACTIVATION ANALYSIS WITH THE COCKCROFT-WALTON

NEUTRON GENERATOR 65

A. Introduction 65

B. Research Activities 67

1. Experimental Spectra of the Elements

and Calculated Sensitivities for 3-MeV

Neutron Activation 67

2. D-D Interference in D-T Analysis. ... 74

vm
TABLE OP CONTENTS (cont.)

Page

3. Efficient Thermalization of 3-MeV

Neutrons. . . 74

C. Analysis of Standard Reference Materials. . 76

1. Analysis of Organometallic Compounds. . 76

2. Analysis of Oxygen in Styrenedivinylbenzene Copolymer 79

3. Analysis of Oxygen in Low-Oxygen

Polyethylene Vials 80

D. Operational Characteristics of the Neutron

Generator 82

E. Health Physics 84

1. Monitoring of Biological Shield .... 84

2. Tritium Contamination and Control ... 85

3. Bremsstrahlung Dose Measurements. ... 87

P. Status of the Neutron Generator Facility. . 89

1. Vacuum Ion Pump Failure 89

2. Ion Source Burnout 89

3. High Voltage Breakdown 92

4. Cooling Problems 93

5. Safety Devices 94

5. ACTIVATION ANALYSIS WITH THE NBS LINAC 96

A. Introduction 96

B. Facilities 96

1. Carbon in High Purity Metals 97

2. Thallium in SRM Glass 101

3. Photofission of Tl , Bi and Pb 102

4. Yttrium in Rare Earths 103

5. Carbon and Oxygen in Sodium 105

6. RADIOCHEMICAL SEPARATIONS 110

A. The Use of Hydrated Antimony Pentoxide

(HAP) for Separation of Elements in

Irradiated Materials 110

1. Introduction 110

IX
TABLE OF CONTENTS (cont.)

Page

2. Experimental Ill

3. Results and Discussion 114

B. The Application of the Antibiotic, Tetracycline, to Chemical Separations 115

1. Introduction 115

2. Experimental 117

3. Results 118

4. Conclusions 121

7. PERSONNEL AND ACTIVITIES

A. Personnel Listing 123

B. Publications 124

C. List of Talks 126

D. News Releases 129

E. References 130

8. ACKNOWLEDGMENTS 133

LIST OF FIGURES

Figure Page

1. Location of Irradiation facilities in the

NBSR 6

2. Irradiation Request Panel 7

3. Send-Receive Terminal - 7/8" system 8

4. Control unit at send-receive station location. 9

5. Penetration of containment building showing

ball valves 11

6. Compressor, ballast tank and central

controller for rabbit system 12

7. Flight lines of armoured polyethylene tubing

for 1 1/2" and 7/8" systems 12

8. Long term irradiation facilities for NBSR. . . 13

9. Temperature sensing labels, NBSR "rabbit" and

quartz blocks used for temperature measurement 16

x
LIST OF FIGURES (cont.)

Figure Page

10. Axial flux distribution in RT-2 and RT-3 ... 17

11. Mounting techniques for 3YY coincidence system

(a) Two phototube technique 23

(b) One phototube technique 23

12. Sample mounting for one phototube technique. . 24

13. Electronics for one phototube technique. ... 25

14

.

Counting room showing shield for Ge(Li)

detector and 4096-channel analyzer 26

15. 47-cc Ge(Li) detector, in shield 27

16. Technique for rapid freezing of ultra pure

reagents

(a) Direct freezing method 31

(b) Polyethylene encapsulation freezing

method 31

17. Gamma-ray spectra of an ancient stained glass

sample at different decay times 34

18. Ether fraction from group separations method . 39

19. Aqueous fraction from group separation

method for SRM steels 40

20. Size distribution of microstandard beads

analyzed 47

21. Gamma-ray spectrum of sodium- free glass

solution 56

22. Resolution of gallium-72 and manganese 56

using a Ge(Li) detector 58

23. Source-detector counting geometry 71

24. Irradiation geometry snowing location of

neutron monitor . . . 72

25. Gamma-ray spectrum from barium irradiated

with 3-MeV neutrons 73

xi
LIST OP FIGURES (cont.)

Figure Page

26. Plan view of experimental configuration used

for D-D neutron thermalization study 75

27. Location of radiation monitoring sites

around the neutron generator facility 84

28. Location of tritium monitoring sites on the

neutron generator 85

29. Bremsstrahlung dose measurement sites around

the neutron generator 87

30. Penning ion source: disassembled 90

31. Penning ion source: inside view of burned

aluminum anode 90

32. Penning ion source: side view of burned

aluminum anode 91

33. Penning ion source: inside view of back

cathode and protruding anode head 91

34. Penning ion source: assembled, showing

mechanical gas lead, and U magnets 91

35. 200kV conductor showing electrical breakdown of insulator 92

36. Target cooling jacket used with Preon 93

37. Beam-side of blank tantalum target showing

burned hole 94

38. Coolant-side of blank tantalum target showing

burned hole. 94

39. LINAC Control Console 96

40. LINAC Activation Analysis Laboratory in the

Radiation Physics Building 97

41. Resistance furnace and collecting train for

carbon determination 98

42. Decay curve for carbon-11 99

xii
LIST OP FIGURES (cont.)

Figure Page

43. Etching of sodium in alcohol-water mixture . . 106

44. Combustion of sodium for carbon determination. 107

45. Coincidence unit for positron annihilation

counting 108

46. Apparatus for separation of oxygen from

sodium 109

47. Sodium removal on a HAP column Ill

48. Structure of the tetracyclines 115

49. Extraction curves with tetracycline 118

50. Extraction of iron and thallium at various

oxidation states 119

51. Extraction with citric acid added. ....... 120

52. Extraction with tartaric acid added 121

53 • Extraction of transition elements with

tetracycline 122

LIST OF TABLES

1. Absolute thermal neutron flux determinations . 17

2. Circumferential flux distribution in NBSR

pneumatic tube facilities 18

3. Cadmium ratios for the NBS pneumatic tube

facilities 19

4. Compilation of temperature .measurements

to date in RT-2 20

5. Compilation of temperature measurements

to date in RT-3 21

6. Results of Mo analysis of W-Mo alloy 28

7. Results of Mo analysis of SRM steels 29

xiii
LIST OF TABLES (cont.)

Table Page

8. Preliminary results using the freezing

technique 33

9. Results of analysis of SRMs lOle, lOlf and

101g 37

10. Results of analysis of SRMs 20g and 36b. ... 38

11. Results of analysis of SRMs 126l, 1262 and

1265 38

12. Results of analysis of SRMs 1155 and l60b . . . 39

13. Comparison of vanadium results in SRM 126l

with and without complement subtraction of

manganese activity 41

14. Results of manganese and vanadium analysis of

SRMs 1261, 1265, 1155, l60b, 20g and 36b .. . 42

15. Sodium content of microstandard beads 48

16. Statistical variation of sodium content with

microstandard size 49

17. Analytical conditions for the determination

of CI, Al and Na in clinical standards .... 50

18. Impurity content of clinical SRMs 50

19. Results of gold analysis in SRM cobaltaluminum flux wires 52

20. Homogeneity of gold in the cobalt-aluminum

flux wires 54

21. Results of analysis of base glass, ppm • • • • 57

22. Results of W analysis of W-Mo alloy 59

23. Results of analysis for selenium 62

24. Selenium content of USPHS samples 64

25. 3-MeV neutron activation data and experimental

sensitivity 70

26. Analysis of organometallic compounds 77

xiv
LIST OF TABLES (cont.)

Table Page

27. Analysis of oxygen in styrene-divinylbenzene

copolymer , 80

28. Analysis of oxygen in low-oxygen polyethylene

irradiation capsules 8]

29. Effect of acceleration voltage on diameter of

beam spot 83

30. Effect of focus control at 200 kV on diameter

of beam spot 83

31. Tritium contamination of the neutron

generator after 30 mA-hrs . of beam operation

on four 0.8 Ci/cm tritium targets (^ 1 1/2

years operation) 86

32. Bremsstrahlung dose measurements as a function of beam current (hydrogen on blank

Cu target) at 200 kV acceleration voltage. . . 88

33. Effect of etching on apparent carbon concentration 100

34. Results of yttrium determination for some

rare earth oxides 104

35. Oxygen in sodium results 109

xv

ACTIVATION ANALYSIS: SUMMARY OF ACTIVITIES

July 1968 to June 1969

Edited by Philip D. LaFleur

The design and Installation of irradiation facilities for

the NBS Reactor are described, and studies of the operating

characteristics of the pneumatic tube facilities have begun.

Procedures have been developed for determining molybdenum in the presence of large amounts of tungsten and

for the determination of uranium. Methods for determining

a contamination free blank for application to liquid samples,

a study of the application of neutron activation to the

analysis of stained glass were begun.

The application of thermal and fast neutron activation

analysis and photonuclear activation analysis to the NBS

Standard Reference Material program is discussed.

Studies of the application of 3-MeV neutrons from a Cockcroft-Walton neutron generator to activation analysis

and of the 3-MeV neutron buildup in 14-MeV neutron activation

has been studied.

The determination of carbon by photonuclear activation

has been .extended to various metals other than sodium,

thallium has been determined in glass samples by photon

bombardment and the photofission of thallium, bismuth and

lead has been studied.

Hydrated antimony pentoxide has been used to separate

sodium from glass samples for subsequent instrumental activation analysis, and the antibiotic, tetracycline, has been

applied to radiochemical separations.

KEY WORDS: Activation analysis; ancient stained glass;

carbon; Cockcroft-Walton neutron generator; coincidence

spectrometer; group separations; homogeneity testing;

hydrated antimony pentoxide; NBS LINAC; NBS Reactor; neutron

flux measurements; photofission; Standard Reference Materials;

tetracycline

.
1. INTRODUCTION

In October, 1968, the Third International Conference on

Modern Trends in Activation Analysis was held at the National

Bureau of Standards. The Conference was chaired by Dr. James

R. DeVoe of the Radiochemical Analysis section with Dr. P. D.

LaFleur as assistant chairman. Over 400 scientists attended

the Conference, including 100 from outside the United States,

making it truly an international conference. Over 160 technical papers were presented, covering all aspects of activation analysis. At the Conference, the report Activation

Analysis: a Bibliography was distributed to attendees. The

report, which was authored by G. J. Lutz, R. S. Maddock,

R. J. Boreni and W. W. Meinke, has been received enthusiastically by workers in the field. This report, which will be

updated periodically, represents the culmination of many

years of effort in cataloging reports and papers in activation

analysis. The Proceedings of the Conference will be published

in June, 1969.

During the past year the application of activation analysis in the NBS Standard Reference Material program has

expanded in the number of elements for which we analyze and

in terms of the sophistication of the analyses

.

Through the use of high-resolution gamma ray spectroscopy and group and/or individual element separations, we

are beginning to approach the capability of performing rapid,

economical multi-element analyses. The most serious problem

facing our Section during the next year is the attainment of

a much higher degree of automation in the processing of samples

Attaining proficiency in automation will require much of our

effort in the coming year.

The NBS Reactor has now operated at power levels from 2

to 10 MW since January, 1969. During this time we have been

gaining experience with the reactor. Unsuspected leaks in

the pneumatic tube systems, have kept us from using all of

the tubes originally installed in the reactor, since Ar

2
activities were above the maximum permissible levels. The

leaks in two of the tubes have been temporarily sealed and

we have been using those tubes. The Reactor Radiation

Division will replace all of the flight lines and air lines

in the existing system with polyethylene tubing similar to

that in the systems installed in the warm laboratories. This

is scheduled for completion during July, 1969.

We have now completed the necessary temperature experiments, as described in the section on facilities, and will be

installing terminals for the warm laboratory pneumatic tube

system during July and August, 1969.

Studies of the solution of blank problems in oxygen analysis with the neutron generator have been completed. A

comprehensive study of the application of 3 MeV neutrons in

activation analysis is nearing completion, and a study of

3 MeV neutron interference in 14-MeV neutron activation has

been studied.

The photonuclear analysis program has continued to operate

whenever time was available. During the year the LINAC began

operating on a 24-hour day, 4 days a week. The activation

project has been operating on 12 hour shifts about three

times a month. Efforts have been channeled in the analysis

of carbon in steels and other metals, the analysis of oxygen

in sodium metal, the determination of thallium in glass, and

the determination of Bi by photofission

.
2. LIAISON AND GENERAL FACILITIES IN THE NBSR

A

.

Introduction

The methods, techniques, and materials employed over the

years in the construction and use of pneumatic transfer systems

in research and test reactors have become generally accepted

without much question and with a minimum of improvement in

the state of the art. Except for a limited number of special

application items, such as rabbit materials, flight tube materials, installation techniques, irradiation terminals, and

instrumentation requirements have remained practically unchanged. However, in the case of a research reactor, such as

14 the NBSR, with a power level of 10 MW, a neutron flux of 10

-2 -1

n.cm . s or higher, and gamma heating rates which could

exceed 1.5 watts/gm in the sample, these standard systems

are no longer adequate. Specifically, a more sophisticated

safety system which will automatically return samples when

the temperature exceeds a pre-determined level, or when there

is a system malfunction or a loss of power, must be incorporated . In addition to our new pneumatic transfer irradiation

terminals, plans are completed for the installation of two

vertical manual facilities and one terminal has been installed.

We feel that the overall system design for use with the NBSR

has advanced the state of the art considerably In the areas

of rabbit system materials, design, safety features and instrumentation. Considerable progress in our flux and temperature

measurement program has been made. Recent data, using temperature sensors, given in this report are being used to predict

safe limits for sample irradiations.

B

.

Liaison

With the advent of full power operation of the NBSR, it

is anticipated that the liaison activities of the section with

outside users of the facilities will increase dramatically

during the next several months. As has been pointed out in

4
previous reports, the NBSR is designed to be a very useful

facility for activation analysis as well as for experiments

in physics and engineering. Our initial liaison activity has

been with Mrs. Jacqueline Olin of the Smithsonian Institution

in Washington, D. C. Since Mrs. Olin did not have at her

immediate disposal the necessary laboratory space or radiation

detection equipment, she has become a guest worker with the

group and uses our equipment. The type of cooperation between

the Activation Analysis Section at the National Bureau of

Standards and outside users can vary from intimate involvement

with the problem, as has been the case with Mrs. Olin, to such

purely mechanical matters such as assistance in packaging

samples for irradiations, assisting in irradiations, explaining operating philosophies of the Activation Analysis Section

and the Reactor Radiations Division and all points in between.

As is pointed out in succeeding sections of this report, we

are obtaining a great deal of useful operating data about the

NBSR which allows us to be of considerable assistance to outside users of the NBSR. We have recently registered our facilities with the Division of Nuclear Chemistry and Technology of

the American Chemical Society in their Major Nuclear Centers

program and welcome inquiries for the use of the NBS facilities,

either on a cooperative basis where the scientists at the

National Bureau of Standards become deeply involved scientifically, or for assistance In using the facilities without this

level of involvement

.

C . Pneumatic Tube System

A new pneumatic tube system has been installed by Reactor

Experiments, Inc. of San Carlos, California* for use in the

NBSR. The specifications for this system were tailored specifically to meet the operating and safety requirements of the

NBSR. In the new system, samples can be sent pneumatically to

three separate irradiation terminals. One of these terminals,

*See disclaimer in Preface.

5
located in the thermal column, will be composed entirely of

graphite in order to minimize flux perturbations in the thermal

column itself. This terminal will accept a standard 1 5/16 in.

11 -? -1 diameter rabbit in a thermal flux of about 3x10 n.cm ~s. . This terminal has three separate send/receive locations, one

in Room C002 in the Reactor sub-basement, and the other two in

Rooms B121 and Bl44 of the warm and semi-warm laboratory area.

The remaining two terminals have vertical access to the

reactor core and will accept a 3/4 in. diameter rabbit. Each

tube has five separate send/receive locations. Although the

design of these terminals has essentially been completed,

tests are being conducted at this time in order to gain a better evaluation of the cooling requirements at their specific

locations in the core. The location of the three new tubes,

plus the relative locations of the four horizontal tubes

already in operation and also the two manual access tubes

discussed below, are shown in figure 1.

RT-4 MT-I

RT-I to4 Horizontal Pneumatic Tubes.

RT-V— Vertical Pneumatic Tube. RT-C Verticol Pneumatic Converter Tube. RT-T Thermal Column Pneumatic Tube.

MT-I 1I4" Manual Tube. MT-2 7/8

" Manual Tube. Figure 1. Location of irradiation facilities in the NBSR

6
1 . Instrumentation and Operation

It was felt that with the possible simultaneous use of

seven pneumatically operated Irradiation locations which could

be accessed from 21 separate send/receive stations, a completely

automated and interlocked system was desirable. Since each

irradiation location has a time limitation on a sample irradiation due to the heat build up, and since a positive seal

is required where the tubes exit from the Reactor Building,

several safety systems have been incorporated. In addition,

to give the reactor operating personnel in the control room the

final decision on a given irradiation, an IRQ, or irradiation

request panel, was Installed. This panel (fig. 2) will provide

information on the status of any tube at any given time. The

tubes are connected to turret type send/receive units pictured

in figure 3

•

Figure 2. Irradiation request panel
Figure 3. Send-Receive Terminal - 7/8" system

A typical control unit of the three-station 1 1/2 in.

diameter tube system is shown in figure 4 . Prom here the

operator, by turning on the power to the unit, locks out all

other stations in the system, (two, in this case). At the

same time he receives a report on the status of the C0 ? propelling system, and the high pressure air system which operates

the valves and switches in the pneumatic circuit. If the

pressure in either of these two systems is insufficient to

provide reliable operation, an indication is given and the

system will not operate. Next, the operator must decide

whether or not the C0

p propelling gas is to be maintained at

a reduced pressure, for cooling, after the sample reaches the

terminal position. The sample return location is next selected

by pressing the appropriate button and then a decision must

be made whether to time the irradiation manually or to use

the timer in the control unit. (As a safety precaution, the

maximum length of time for which a sample can be irradiated in

each location is predetermined and programmed into each system.

8
Figure 4. Control unit at send-receive station location

Therefore, regardless of which type of timing is selected, the

sample will automatically return if this maximum irradiation

time is reached.) This procedure completes the programming,

and the system is ready to send the sample. To do this the

permission Granted/Requested button is pressed. The indicator

light comes on and blinks on the control panel and on the IRQ

panel in the control room. If the reactor operator is ready

for the sample to be sent, he presses the corresponding button

on his IRQ panel, granting permission for the irradiation to

take place. The sample may now be dispatched by activating

the Send Sample button on the station control panel. As the

sample approaches the core, a photocell system is triggered

and indicates on the IRQ and station control panels the arrival

of the sample in the reactor core. Although the sample travels

at a speed of around 50 feet/second, the impact is reduced by

the use of an air-cushion braking system.
While the sample is in transit, entering or leaving the

core, all other stations and systems are locked out and will

not operate until the routing sequence has been completed.

The return of the sample at the end of an irradiation is accomplished by the automatic timer or by pressing the Manual

Return button.

2 . Safety Systems

As mentioned above, both the air pressure for operating

the valves and switches in the system, and the C0

p propelling

gas pressure must be above a certain level before the system

can operate. The system also has a Stuck Capsule alarm. This

is accomplished by a delay timer which has been pre-set for

some time greater than that required to return a heavy capsule.

As the capsule return cycle is actuated, the timer starts and

is reset as the capsule arrives at the receiver station. If,

however, the pre-set time is exceeded, the system assumes a

"stuck" or "lost" capsule situation' and initiates the Stuck

Capsule alarm. This alarm turns on lights at the station

control and IRQ panel, and activates three rotating beacon

lamps at various places in the building where personnel could

be near the tubing and thus be exposed to radiation from the

stuck capsule.

Another safety feature is the Automatic Emergency Return

System (AERS). This system is actuated automatically in the

event of a major scram, electrical failure, or inadvertant

loss of C0

p pressure. It can also be activated manually by

pressing the Emergency Return button on the station control

panel. The AERS sequence operates as follows. When initiated,

all previously programmed logic is isolated, the system is

switched to emergency power, and an external CO^ bottle fires

out any and all rabbits, in sequence, to a separate, shielded

location. In the event of a major scram, this sequence is

preceded by the closing of the three ball valves on either side

of the containment wall (fig. 5) thus preserving the integrity

of the reactor containment.

10
.-' Figure 5. Penetration of containment building showing ball

valves

The compressor and ballast tank which supply the recirculating supply of COp for propelling the samples, and the central relay panel for the system are shown in figure 6.

The flight tubes (fig. 7) consist of a 1/8 in. thick polyethylene inner sleeve, through which the rabbit travels; a steel

armour covering for structural reinforcement, and a polyvinyl

chloride covering over the armour which provides a corrosionresistant finish.

D. Manual Irradiation Facility

The manual irradiation tubes allow irradiations for a

longer period of time than would be possible in a rabbit tube.

The two types of manual irradiation tubes for installation in

the NBSR are shown in figure 8. One type, holding a single

sample container at a time, utilizes 1-1/2 in. diameter graph14 -2 -1 ite buckets in a flux of ^10 n.cm .s . These buckets will

be suspended at the end of a thin aluminum wire. The second

11
Figure 6. Compressor, ballast tank and central controller

for rabbit system

m%*

Figure 7.. Flight lines of armoured polyethylene tubing for

1 1/2" and 7/8" systems

12
-D20 LEVEL LEVEL SENSORS

INLET DpO

SAMPLE GUIDE

.HI-DENSITY LEVEL SENSORS CONCRETE

Figure 8. Long term irradiation facilities for NBSR

type will permit irradiation of up to seven 3/^ in. diameter

samples simultaneously. This system will be located in the

13 -2 -1 reflector area where a flux of about 5x10 n.cm . s will be

available

.

The first system .has been installed and temperature studies

indicate that with no circulation of the DpO in the tube, the

10 MW equilibrium temperature of l80°P can be expected. However, when the DpO is circulating, the temperature is only

slightly above ambient. Irradiation times equivalent to one

reactor cycle Ol8 days) will be possible.

E. Status of Existing 1 1/2 in. Diameter Horizontal Pneumatic

Tube Systems

The existing four 1 1/2 in. horizontal rabbit tubes have

been discussed in detail in a previous report [1]. During

the coming year the tubing in this system will be changed

from aluminum to the armoured polyethylene type used in the

new system since it was found that the many joints in the

aluminum tubing resulted in excessive leakage of the C0 9

.

13
p , Progress of Irradiation Facilities in NBSR

Insofar as the pneumatic tubes are concerned, all construction and installation of flight tubes, send/receive stations,

main IRQ panel and safety systems have been completed. The

irradiation terminals are in the design stage. These designs

will be strongly influenced by experimental results of temperature and irradiation dose studies presently being conducted

using the manual vertical facility and the existing horizontal

tubes. The graphite terminal to be installed in the graphite

thermal column has been completed. This terminal will be

installed as soon as the shield plug to be located above the

terminal has been suitably modified. Considerable progress

has been made on the long term vertical manual facility. The

final design was recently approved and the terminal has now

been located in the reactor core and is undergoing tests.

A temporary flow system has been set up to circulate D

p

in

the terminal for 10 MW reactor operation. Experience with the

temporary flow system will be used for a permanent installation

incorporating all necessary safety features. A lead cask has

been constructed to facilitate safe removal of samples from

this terminal.

A complete modification of the existing horizontal pneumatic tube facility is being made by the Reactor Operations

Group. It is hoped that all four tubes will be in service in

the near future for operation at 10 MW reactor power.

(F. A. Lundgren, S. S. Nargolwalla, P. D. LaFleur)

G. Measurement of Temperature and Neutron Flux Characteristics of the NBSR Pneumatic Tube Irradiation Facilities

RT-2 and RT-3

As discussed earlier in this report, the NBSR went critical

and became available for use for the first time this past year.

In order to fully utilize the capabilities of this reactor for

neutron activation analysis, information must be known about

each of the irradiation facilities available. The information

14
required pertains to the thermal and fast neutron flux distributions and the temperature characteristics of different

materials irradiated within each facility. These measurements

are in progress. This segment reports on the information

obtained to date on the pneumatic tube irradiation positions

RT-2 and RT-3.

1

.

Design Information on Pnemuatic Tubes RT-2 and RT-3 . Specific information on these irradiation positions has

been previously discussed [1,2]. Briefly, RT-2 and RT-3

are both located horizontally in the reactor core area. Pneumatic tube RT-2 terminates inside the core area itself while

RT-3 terminates in the D p moderator just outside the fuel

gap at midplane. At the outset of these measurements, very

little was known about the temperature characteristics of

various types of materials within these irradiation positions,

or even whether polyethylene rabbits could be used for more

than a few minutes.

2

.

Experimental

a. Neutron Flux Measurements . The neutron flux measurements were of two types. First, the homogeneity of flux

distribution within the rabbits was measured for both RT-3

and RT-3. These measurements were made using 5 mil copper

foil distributed both circumferentially and axially throughout

the rabbit. In most cases the copper foils were gross gamma

counted shortly after irradiation, and the flux distribution

calculated. Second, the neutron energy distribution throughout the rabbits was measured. These measurements were of two

kinds - cadmium ratio determinations and threshold foil determinations. The cadmium ratio determinations (with both gold

and copper) have been completed and the threshold foil measurements are in progress.

b. Temperature Measurements . The temperature measurements

were made using small temperature indicating stickers attached

to each sample. These stickers were obtained in various temperature ranges, each indicator stating the temperature at

15
which it irreversibly changes from white to black. These

stickers were attached to various types of materials (polyethylene, quartz and lead) as shown in figure 9. A series of

irradiations were made starting with small quantities of samples at low power levels and slowly increasing the irradiation

times, the sample size, and the power level with each irradiation. To date, information has been obtained through the

5 MW level for both irradiation positions RT-2 and RT-3 using

the three materials mentioned in various configurations.

»'...MiW T^|||fflH¥r

I; w»

/ONlTOBg;

Figure 9. Temperature sensing labels, NBSR "rabbit" and

quartz blocks used for temperature measurement

3* Results

a. Neutron Flux Measurements . The results of these

measurements are found in tables 1-3 and figure 10. Table 1

shows the absolute thermal neutron flux as determined by both

gold (0.0005 in. thick) and copper. These values are compared to the design estimates as given in NBSR-10 [2] . There

16
1.32 V 1 1 1 1 1 1 1 1

1.24 ^ -

1.16 Q-^—^____^ RT-2 —

X

_)

U. - -

NEUTRON b00 -

RELATIVE bo -

0.92 —

""*"^-*.^ RT-3

0.84 - ^^v ^

0.76 1 1 1 1 1 1

1

1 2 3 4 5 6 7 DISTANCE FROM PROXIMAL END OF RABBIT (cm)

Figure 10. Axial flux distribution in RT-2 and RT-3

Table 1. Absolute Thermal Neutron Flux Determinations

Thermal Neutron Flux

(n.cm . s ) Measurement

Calculated

Experimental

Experimental

Element

Gold

Copper

RT-2

7.3xl0 13

6.9xl0 13

5.3xl0 13

RT-3

9.4xl0 13

,13 5.9x10"

4.8x10 13

Obtained from NBS Report 9081 (NBSR-10)

Contribution to foil activity; measured at proximal end.

17
is a significant difference in the experimental flux as determined by gold and by copper. This discrepancy most likely

results from the much greater variation of the gold cross

section versus neutron energy compared to copper. Results

for both elements show that RT-2 has a slightly higher thermal

(subcadmium; <0.l4eV) neutron flux than RT-3, a fact not expected from design estimates.

The circumferential flux distribution in the rabbits as

determined by copper foils is given in table 2. Since the

incore rabbit terminals are horizontal, the location of the

foils within the rabbit itself was measured from the end

closest to the core. The results indicate no significant flux

variation circumferentially in either pneumatic tube. However,

as shown in figure 10, the axial neutron flux distribution does

vary significantly. This axial variation is quite irregular

for RT-2. However, for RT-3 the axial variation is almost

linear over the entire length of the rabbit, falling off

slightly at the distal end.

Table 2. Circumferential Flux Distribution in NB.SR

Pneumatic Tube Facilities

Pneumatic Location Relative No. of

Tube Facility in Rabbit Neutron Flux Monitors

RT-3 Proximal 1.000±0.004 a

7

RT-3 Middle 0.928±0.003 6

RT-3 Distal 0.829±0.003 7

RT-2 Proximal 1.300±0.007 7

RT-2 Middle 1.172±0.002 6

RT-2 Distal 1.136±0.003 7

Uncertainties are ts//n at 95$ confidence level.
The results of cadmium ratio determinations made at both

ends of the rabbit for RT-2 and RT-3 using copper and gold

foils with 20 mil cadmium covers [3] are shown in table 3.

The cadmium ratios for RT-2 were approximately 4 for gold and

16 for copper, and were the same at both ends of the rabbit.

This large fast neutron component is to be expected due to the

position of RT-2 between the fuel elements. Tube RT-3 has

much larger cadmium ratios for both gold and copper, and shows

a significant difference in cadmium ratio between the proximal

and distal ends of the rabbit. This is also expected since

the RT-3 position extends further into the D

p moderator. It

is obvious that when a sample irradiation position is chosen,

consideration must be made for the fast neutron component in

the neutron flux.

Table 3. Cadmium Ratios for the NBS Pneumatic Tube

Facilities

Gold-Cadmium Ratio Copper-Cadmium Ratio

Pneumatic Location

Facility in Rabbit Au(Cd ) Average Cu(Cd ) Average

RT-3 Proximal 9.6 \ 10.6 65.0 I 65.3

11.6 / 65.5 /

RT-3 Distal 15.7 X 16.5 80.5

17.2 / RT-2 Proximal 3.92 - 16.0 \ 16.0

15.9 J RT-2 Distal 4.03 - 15.8

19
b. Temperature Measurements . The compiled results of temperature measurements made to date are shown in tables 4 and 5.

They are self-explanatory. The results listed are for maximum

values only. A large number of preliminary irradiations were

made to obtain these values. Much additional work will be

performed at the 10 MW power level to precisely determine the

safe limits for sample irradiations.

Table 4. Compilation of

to Date

Temperature Measurements

in RT-2

Wt

Material Configuration g

Power Length

Level of

MW Irrad. Results

Polyethylene

Polyethylene

Polyethylene

Polyethylene

Quartz

Quartz

Quartz

Quartz

Quartz

Quartz

Quartz

Lead

Sheet, 2 mm

thick 1

Sheet, 2 mm

thick 1

Empty rabbit

Empty rabbit

7 mm dia tubing

open ends 1

Solid block 14

7 mm dia tubing

open ends 1 3

7 mm dia tubing

open ends 1 5

Solid block 4. 6 5

7 mm dia tubing

open ends 3. 5 5

7 mm dia tubing

open ends 3. 5 5

Sheet, 1 mm

thick 1 2

20 m

20 m

10 h

20 h

60 m

60 m

80 m

60 m

20 m

20 m

20 m

120±5 °F

130±5 °F

105±5 °.F, no

significant

radiation

damage

Small amount

radiation damage;

still usable for

irradiations

.

120±5 °F

155±5 °F

125±5 °F

125±5 °F

210±10 °F

145±15 °F

200±10 °F

5 m 200120 °F

20
Table 5. Compilation of Temperature Measurements

to Date in RT-3

Material

Length of

Power IrradiWt Level ation

Configuration g_ MW Minutes Results

Polyethylene Sheet, 2 mm

thick 12 <100 °F

Quartz

Quartz

Quartz

Quartz

Solid block 15 120

7 mm dia tubing

open ends 1

Solid block

Solid block

3

7 5

14 5

105± °F

20 <100 °F

20 130±10 °F

20 155±5 °F

Lead

Lead

Lead

Sheet, 1 mm

thick

Sheet, 1 mm

thick

Sheet, 1 mm

thick 2.1

30

20

20

120±10 °F

160±10 °F

180±10 °F

Lead-Quartz Solid strip of

lead 1.13

(Pd)

Sheet in sealed

quartz ampoule 3.63

(Si0

2

)

50 155±5 °F

(D. A. Becker)

21
3. ACTIVATION ANALYSIS WITH A NUCLEAR REACTOR

A

.

Introduction

During this year we have continued our efforts to attain

a high degree of versatility in terms of numbers of elements

analyzed and matrices, while maintaining the highest possible

standards of accuracy and precision. We intend to continue

these efforts during the coming year, as well as begin investigations into areas in which we have not been active, such as

epithermal neutron activation analysis and automated activation analysis. Now that irradiation facilities are available

at the National Bureau of Standards, we shall also increase

our efforts in activation analysis using short-lived isotopes.

Operation of the NBSR has been a great advantage to the

Reactor Activation Analysis group as we are now able to perform all of our work at the Bureau of Standards. We have,

of course, spent a considerable amount of time becoming acquainted with the operating characteristics of the reactor,

and with the safety criteria for performing irradiations.

As usual, much of our research has been involved with the

analysis of NBS Standard Reference Materials, but as our

expertise improves, our research effort is becoming less

applied to specific samples.

B. Facilities

1. Modifications to a Bet a- Gamma- Gamma Sum Coincidence

Spectrometer

This $YY sum coincidence spectrometer has been described

previously [3]. As originally designed, the sample to be

counted was sealed between two thin plastic scintillator discs

and optically coupled to two 3/4 in. photomultiplier tubes.

(fig. na). A pulse from both photomultiplier tubes was required to indicate an event in the plastic scintillator, thus

effectively eliminating noise pulses from the photomultiplier

tubes. This feature did lower the background ^10-20$ (almost

entirely in the low energy region [3]), however, we recently

22
PLASTIC SCINTILLATOR

S. S. TUBE 3/4" DIA. PLUGS, MOUNTED

"TO PHOTOTUBE

(a)

OPTICAL COUPLING WITH

SILICONE GREASE

SAMPLE

BLACK PLASTIC TAPE

OPTICAL COUPLING WITH

SILICONE GREASE

zLIGHT-TIGHT

PLASTIC CAP

PLASTIC SCINTILLATOR

(b)

Figure 11.

PLASTIC HOUSING MHV

Mounting techniques for 3yy coincidence system

(a) Two phototube technique

(b) One phototube technique

discovered inconsistencies in counting efficiency. Repetative

fid counting of the same Co standard after remounting gave

results differing by as much as 50% . After extensive testing

and elimination of most of the possible electronic malfunctions, the remaining possibility was the optical coupling

of the scintillator disks to the phototubes. Since the phototubes had ^1 mm play from side to side in the stainless steel

tubes, the possibility existed that one of the phototubes

might drop away from the scintillator disks during counting

resulting in loss of optical coupling. This would account

for the variable results in the Co standard counts..

To eliminate this problem, a revised detector system was

devised using only one phototube (figs, lib & 12). In this

new system a 3/4-in. photomultiplier tube with associated

electronics (fig. 13) was potted in a 1 in. OD bakelite tube

with black RTV silicon compound. The junction plug which was

23
Figure 12. Sample mounting for one phototube technique

outside the phototube housing before, is now inside the bakelite housing. The HV and signal leads are brought through the

rear of the housing using an MHV connector. A tight fitting

cap of bakelite was made to enclose the source. Black electrical tape is used to keep the cap in place and prevent anylight leaks

.

The new system has completely eliminated the counting

inconsistencies. Repetative counting of both cobalt-60 and

sodium-24 standards gave the following relative efficiencies:

Tube A=91. 2*±2.1$; Tube BE100$; tube C = 9^ . 3±2 . 2% . The averages

quoted are for three separate determinations, for each detector,

with 10,000-50,000 counts per determination. This system

was used for the analysis of picogram quantities of sodium

in microstandard particles, as described in part 3-

24
a

<*

T<10 J

1 'llVr 1"! ANODE

DYNODE

CATHODE

R, = 0.39M I/2W

R2 _R|| = 0.2MI/2W

R,2

= .IMI/2W

C, = 260pf500V

C2

= 60 pf 500V

C3

= .Ol/if 2kV

R| 2 —VW

~7L

Figure 13. Electronics for one phototube technique

(D. A. Becker)

2 . Counting Room

During the year we acquired a new 47-xx Ge(Li) detector

with a system resolution of 4 keV. A view of the counting

room configuration to include the shield for the new detector

is shown in figure 14. Also shown in the picture is the new

high-speed point plotter that was acquired during the year

for plotting gamma-ray spectra. A view into the Ge(Li) detector shield is shown in figure 15.

25
Figure 14. Counting room showing shield for Ge(Li) detector

and 4096-channel analyzer

C . Research Activities

1. Activation Analysis for Molybdenum in the Presence

of Large Amounts of Tungsten

As part of the certification of SRM 480, a W-Mo alloy

designed to be used as an electron microprobe standard, the

Section was asked to determine Mo in some samples of the alloy.

The samples weighed about 5 mg each and contained nominally

80% W - 20% Mo. Tungsten could easily be determined in the

samples nondestructively , but a chemical separation was required for the Mo determination because of interference with

9 9m 99 the 0.142 MeV y-ray of yy Tc used as a measure of the ^Mo

1 Pi 7

activity by the much more intense 0.134 MeV Y_ray of W.

26
Figure 15. 47-cc GeCLi) detector in shield

In steels, where the W-Mo ratio is of ten much lower than

in these samples, it is frequently possible to allow the

W to decay until this interference is negligible [1,4]

However, in these samples the W-Mo ratio is much too high to

permit this approach and so a chemical separation was developed.

The method chosen was based on the results of research

carried out in this group on the extraction behavior of a number of metal ions with bis (2-ethylhexyl) orthophosphoric

acid (HDEHP) [5,6]. Using 11 M HNO- as the aqueous phase

Mo can be quantitatively extracted into HDEHP while Tc is

not extracted. The analytical procedure used incorporates

these facts by first extracting the Mo into HDEHP from 11 M

27
HNO-. Tungsten is also extracted. Then, after allowing time

3qqrn qq

for Tc to equilibrate with its Mo parent, the organic

QQm phase is washed with 11 M HNO- . The ^"TTc is quantitatively

removed from the organic phase and thereby separated from

both W and Mo. This wash solution is counted directly using

a 3 in. x 3 in. Nal(Tl) detector in conjunction with a 400

99m channel pulse height analyzer and the Tc peak used as a

measure of the Mo present in the sample. The method is

simple and rapid. Excellent decontamination from virtually

all interfering activities is obtained. Corrections are

ma<

99,

99m made for decay of Tc after separation and for decay of

#1

#2

#3

#4

Mo up to the time of separation. No determinations of

chemical yield are required since the yield in each step is

quantitative. Four of the small samples of the W-Mo alloy

were analyzed and the following results were obtained.

Table 6. Results of Mo Analysis of W-Mo Alloy

Sample % Mo

23.82±0.5^ a

24.24±0.87

24.36±0.51

23.4l±0.42

Limits quoted are ts//n for the 95$ confidence level;

n = 6.

Analysis of similar samples by differential spectrophotometry [7] gave results of 23.90±0.06# and 23.75±0.15#

which is in good agreement with the activation analysis results.

The procedure has also been applied to the determination

of Mo in NBS Standard Reference Material steels with the

following results

28
SRM No.

lOle

20g

36b

Table 7. Results or Mo Analysis of SRM Steels

% Mo

Found In this Work NBS Certified Value

0.428 ±0.005 0,426

0.0026010.00009 0.002, 0.003

1.08 ±0.02 1.00

A paper describing this work was presented at the 156th

National Meeting of the American Chemical Society held at

Atlantic City, New Jersey, September, 1968. This paper has

now been submitted for publication.

2 . Ultrapure Reagents: Use of the Freezing Technique

in the Quest for the Absolute Blank

In any analysis procedure, it is imperative to know the

"blank" value, that is, the contribution to the analysis by

the technique itself. Ideally, the blank will be zero, as

when in activation analysis a solid is irradiated and then

etched to remove any surface contamination. The sample can

then be weighed and the analysis continued without consideration of a blank value. However, when sub-ppm levels of trace

elements are to be determined in liquids, the possibility of

leaching of impurities from the container is a serious problem,

Brune [8] has shown that significant amounts of impurities

are picked up from either polyethylene or quartz by both

leaching and nuclear recoil processes. In the analysis of

ultrapure reagents at NBS this blank problem is especially

critical, since these reagents will be used by many other

analytical competences for trace element analysis, and the

"true" impurity content must be known.

The ultrapure reagent program and the special problems

associated with it has been discussed previously [9]. This

preliminary work demonstrated the need to develop a method of

determining the blank values for the analysis of liquid

29
reagents. An alternative would be the development of a procedure for analysis of these reagents which would eliminate

contamination by the container, thus producing an "absolute

blank."

The technique of freezing liquid samples prior to irradiation has been used previously both for the retention of

volatile elements and for the prevention of contamination from

container materials [10,11], The procedure used consisted of

irradiating the frozen samples allowing the surface to melt

slightly to remove surface contamination, and processing for

radiochemical separation or counting. After consideration of

the possible methods of blank determination and reduction,

this freezing technique was chosen as the most promising.

This report describes experiments performed in an attempt to

use the freezing technique to determine the trace metals

content of high purity water without any contamination from

the container materials. (This could then be considered an

absolute "blank" for any reagents prepared with this water.)

a. Experimental

There has been a great deal of thought on which method

is best for the taking of the sample. The high purity water

used in the procedure development was filtered, deionized,

l8Mft water from a recirculating still. Ideally the sample

should be frozen without contacting any container after removal from the still. This was accomplished by letting the

water drip slowly from the top directly into a clean dewar of

filtered liquid nitrogen (fig. 16a). The ice is caught in a

clean platinum gauze basket before touching the sides or

bottom of the dewar [technique 1] . There was, however a

possibility that the droplets might pick up microscopic bits

of frozen water from the liquid nitrogen so a second technique

was developed to cross-check the results. This consisted of

allowing the water droplets to fall directly into a frozen

section of rigorously precleaned medical grade polyethylene

30
PVC TAP FROM

RECIRCULATING

PURIFICATION SYSTEM

FROZEN WATER

DROPLETS

DEWAR

PLATINUM GAUZE

BASKET

LIQUID N2

CLEAN POLYETHYLENE

TUBING

DEWAR

FROZEN WATER

SAMPLE

LIQUID N2

Figure 16. Technique for rapid freezing of ultra pure

reagents

(a) Direct freezing method

(b) Polyethylene encapsulation freezing method

tubing. The bottom end of the tubing was heat sealed and

immersed in liquid nitrogen leaving ^1 inch protruding (fig. 16b).

The water drops then only contact the polyethylene at -196°C,

at which temperatures they freeze immediately [technique 2]

.

Using this technique the samples are larger and more easily

"etched" after irradiation.

The samples are sealed before irradiation to prevent contaminants from contacting the samples, and to prevent atmospheric water from freezing on the surface of the samples and

contaminating them. Using technique 1, the individual frozen

droplets must be selected and transferred to cooled polyethylene tubing similar to the one described for technique 2.

Forceps with polyethylene tubing covering the tips were used.

For both techniques the tubing was clamped shut with a hemostat and the end heat sealed. This can be done with the

other end either in liquid nitrogen, or quickly lifted out

during sealing and then reimmersed. The samples are then

ready for irradiation.

31
To date all irradiations were in the NRL Reactor, and were

accomplished using a polyethylene bucket filled with dry ice.

Refilling the bucket was necessary at about 15 minute intervals. After irradiation for 30-60 minutes the samples are

removed and stored in dry ice for subsequent "etching" and

counting (with or without radiochemical separations).

With the new NBSR facility, the irradiation procedure

should be a great deal simpler. By packing a rabbit full of

dry ice (with vent holes to let the gaseous carbon dioxide

produced escape) an irradiation of at least 5-10 minutes at

i 13 -2 -1 4x10 n.cm . s should be possible without refilling, which

would be roughly equivalent to the 30-60 minutes at NRL.

Longer irradiations, or use of a higher flux pneumatic tube,

could be used to increase the sensitivities, although replenishing the dry ice might be required.

After irradiation, the samples are removed from the polyethylene container, allowed to warm to ^0°C, and "etched" to

remove surface contamination. With the frozen individual

droplets (technique 1) the samples are placed into a clean

platinum gauze basket, and immersed into several different

water baths, sequentially. The water baths are cooled to

approximately 5-10°C, so the samples do not melt too quickly

and the samples are rolled about in the basket to equalize surface contact and "etching." After about 20% of the mass has

been removed, a blast of high pressure nitrogen gas is used to

remove excess water, and the samples transferred to a clean

container and allowed to melt. For samples prepared with

technique 2, the solid sample is removed from the polyethylene

tubing by cutting off one end, inserting halfway in a beaker

of room temperature water for ^5 seconds, then squeezing.'the

bottom of the tubing. The "etching" procedure is identical.

The element (s) of interest determined whether or not

radiochemical separations were used. Sodium-24 activity is

usually present in amounts that may mask other activities so

41 radiochemical separations were usually employed. Also Ar

32
lfi and P produced from dissolved air and the water matrix may

also mask other activities. The samples have been counted

using gamma-ray spectroscopy, almost entirely with (Nal(Tl)

detectors to increase the sensitivity, although some work has

been done with a Ge(Li) detector.

In the future, 3y and $yy coincidence counting will be

used to lower the background.

b . Results and Discussion

Activity to date has been directed primarily to developing

the techniques necessary to give meaningful results. Tests

137 of the etching technique with Cs indicated decontamination

of at least a factor of 10 from contamination on the surface

of the frozen droplets. Later refinements in the technique

should increase this factor.

A test of the technique itself was made using copper which

was found to be present in the high purity water irradiated

as a liquid. The results in table 8 indicate that some

contamination was picked up from the container when the samples

were not frozen.

Table 8. Preliminary Results Using the Freezing

Technique

Copper Sample

Cone. Size

No

.

Sample

1 Liquid Water

Liquid Water

3 Frozen Water

4 Frozen Water

Procedure (ppb) (g)

Poly Tubing 0.74 5

Synthetic Quartz 0.74 5

Technique 1 0.4 0.9

Technique 2 0.45 5

(D. A. Becker)

33
3 . Analysis of Ancient Stained Glasses by Neutron

Activation Analysis

In cooperation with the Smithsonian Institution and Mrs

.

Jacqueline Olin, a guest worker at NBS from the Smithsonian,

a program is in progress to study the characterization of

medieval stained glasses using neutron activation analysis.

Procedures have been developed using stained glass from the

Washington National Cathedral and simulated medieval stained

glasses prepared by Corning Glass, which incorporate both

nondestructive analysis and radiochemical separations. Sodium

removal with hydrogen antimony pentoxide (HAP) as described

by Girardi and Sabbioni [12] and solvent extraction procedures

have been used. To date four blue glasses and one green glass

have been examined and all have different compositions of both

major and minor constituents. Two spectra of one of these

glasses taken 8 min. apart are shown in figure 17- The decay

of 2.3 min. Al, 3-75 min, 52V, and 5-10 min. 9 Cu can be

readily observed. It is expected that these procedures will

be applied to the analysis of medieval stained glass samples

in the near future.

L78 MeV

Figure 17. Gamma-ray spectra of an ancient stained glass

saraDi e at different decay times

(B. A. Thompson and J. Olin)

34
4 . Rapid Radiochemical Separation of Uranium

The Activation Analysis Section is planning the compilation

of a set of standard radiochemical separation procedures for

use in the activation analysis of many of the elements. Many

times when an analysis is needed quickly but cannot be done

nondestructively , much time and effort is expended in searching

out and evaluating the many radiochemical separations in the

literature. It would be advantageous to have a set of standard

procedures, including methods of separation from various

matrices, for a given element. These would have been proven

by repeated use, and would take advantage of the advanced

instrumentation now available. As an example, for most situations, a quantitative separation of 2 to 5 elements in a group

is much to be preferred to the isolation of a single element

in high purity with a 60$ yield, since solid state detectors

usually may be used to resolve the gamma rays of a group of

this size with little or no uncertainty.

An example of how handy this type of compilation would be

occurred recently when the need for an analysis for trace

amounts of uranium was requested. The inorganic chlorides to

be analyzed contained many radioactive isotopes upon neutron

irradiation, and the low energy, 75 keV, gamma-ray of uranium239 was completely lost in the Compton continuum. The possibility of using the decay product neptunium-239 was considered

also, but the matrix did not decay rapidly, and thus both techniques would require a radiochemical separation. For analvsis

239 using U, which has a 23-5 m half-life, the uranium separation would have to be rapid.

Solvent extraction characteristics using the reagent bis(2-

ethylhexyl) orthophosphoric acid (HDEHP) have been studied in

great detail in this laboratory, and it was inevitable that

this reagent should be suggested for the analysis.

35
a

.

Experimental

The samples and standards were irradiated for 30 seconds

in the NBSR RT-3 pneumatic tube facility, at a thermal neutron

13 -2 -1

flux of approximately 1x10 n . cm s . After irradiation,

they were allowed to cool for 1-2 minutes, and then transferred

to a radiochemical hood for processing.

Previous experience with this reagent suggested that

uranium would extract into HDEHP quantitatively in about 10

seconds from solutions of 1-10M nitric or perchloric acids.

The procedure is as follows . The irradiated sample was

dissolved in 10 ml of 8M HN0~ containing ^50 ug of U carrier

in a separatory funnel. Ten milliters of a solution of 0.75M.

HDEHP in petroleum ether was added and the mixture vigorously

shaken for ^1 minute. The two phases were allowed to separate

and the aqueous phase was drawn off and discarded. The organic

phase was washed twice with 10 ml of 8M HNO^ for 30 seconds

each time. After drawing off the last wash, a transfer pipet

was used to very carefully remove the organic layer. This

procedure prevents possible contamination by small amounts of

the aqueous phase still clinging to the sides of the funnel

stem. Total time for analysis is approximately 15 minutes.

b . Results and Discussion

Tracer studies using synthetic samples showed consistent

yields of >99 • 9% • While some fission products are known to

extract using these conditions, their activity was much less

than that of ~ 39U.

Analysis of the samples submitted revealed no detectable

amount of uranium. Sensitivity calculations indicate that the

upper limit of uranium concentration in these samples is 0.016

ppm.

Increased sensitivity could be obtained for the analysis

by irradiating in a higher neutron flux or for a longer period

of time or both. Analysis of sub-ppb levels should be possible

for a 20 minute irradiation at 10 MW, although the extraction

36

II
101 97 of several of the fission products (e.g. Mo and Zr) may

require another decontamination step (e.g. washing with HP may

be possible)

.

(D. A. Becker and P. D. LaPleur)

D. Analysis of Standard Reference Materials (SRM ) 1. Activation Analysis of SRM Steels Using Group Radiochemical Separations

The group radiochemical separation procedure described in

NBS Technical Note 458 [1] and in the literature E^J has been

applied to the analysis of a number of SRM steals. The procedure consists of two solvent extraction steps, thus dividing

the sample into three groups. In the first step W and Mo are

extracted into CHC1~ from 3 N HC1 as the a-benzoinoxime complexes. In the second step Fe and Sb are extracted into isopropyl ether from 7.5 N HC1. The W-Mo fraction is counted

with a Nal(Tl) detector for W and, after ^W has decayed, for

Mo. The two fractions from the ether extraction are counted

with a Ge(Li) detector for Cu, Cr, As, Co, Sb, Ga, and La.

Where appreciable amounts of Ta are present, this element can

be determined in the W-Mo fraction. The results reported on

the various steels analyzed are given in the following tables.

Table 9. JResults of Analysis of SRMs lOle, lOlf, and

101g

Element %_ 101e a lOlf lOlg

0.0289 ±0.00099 0.0292 ±0.0011

0.0864 ±0.0028 0.0901 ±0.0023

0. 00 02 6l± 0.00002

6

0.000282± 0.000023

0.00862 ±0.00077 0.00286 ±0.00042

0.00283 ±0.00016 0.00277 ±0.00017

0.00398 ±0.00028 0.00325 ±0.00009

0.00092 ±0.00013 0.00130 ±0.00015

Limits quoted are ts//n for the 95$ confidence level;

n = 6.

n = 5.

37

Cu 0.355 ±0.0l6 b

Co 0.017 ±0.013

W 0.0458 ±0.083

Mo 0.445 ±0.043

As 0.00769±0. 00022

Ga 0.00444±0. 00046

Sb 0.00504±0. 00054
Table 10. Results of Analysis of SRMs 20g and 36b

Element % 20g 36b

Cu 0.003396±0.00079

a 0.1882 ±0.0018

Co 0.00276 ±0.00024 .01l44±0.000l6

Cr 0.0177 ±0.0013 2.286 ±0.030

Mo 0.00260 ±0.00009 1.08 ±0.02

Limits quoted are ts//n for the 95% confidence level;

n = 6

.

Table 11. Results of Analysis of SRMs 1261, 1262,

1265

1262 1265

0.00530 ±0.00015

0.000307±0".000020

0.00607 ±0.00029

0.000068±0. 000020

0.00600 ±0.00045

<0.02

0.0104 ±0.0004 <0.001

0.00055±0.0004

aLimits quoted are ts//n for the 95% confidence level;

n = 6

b

n = 5.

Element % 1261

Cu 0.0418 ±0.00l4 a

As 0.0174 ±0.0007

Co 0.0253 ±0.0008

W 0.0152 ±0.0009 b

Cr 0.605 ±0.020

Zn - Sb O.OOl45±0.00011b

Ta 0.0277 ±0.0031b

La 0.00077±0. 00004

38
Table 12.

Element %

Cu

As

Co

Results of Analysis of SRMs 1155 and l60b

1155 160b

0.194 ±0.006a

0.0097±0.0006

0.101 ±0.005

0.199 ±0.006

0.0107±0.0004

0.101 ±0.0003

Limits quoted are ts//n for the 95$ confidence level

n = 6

Not all the above elements were determined in each steel and

in fact for some, information was only required about a few

elements. Typical spectra of the two fractions obtained from

the ether extraction are shown in figures 18 and 19.

"55c

>

Gamma Energy , MeV,

Figure 18. Ether fraction from group separations method

39
"(75 cc

0>>

Gamma Energy , MeV.

Figure 19. Aqueous fraction from group separation method

for SRM steels

(B. A. Thompson)

2 . The Determination of Manganese and Vanadium in SRM

Nos. 16b, 115, 1265, 1261, 36b, 20g.

As a part of the NBS Standard Reference Material program,

manganese and vanadium were determined in samples of six different steels using nondestructive neutron activation analysis.

Similar analyses were reported in previous progress reports

[1,9]. Possible interfering contributions to the gamma rays

used in the analysis were evaluated by their half-life determinations and high resolution Ge(Li) gamma ray spectroscopy. No

such contributions were discovered. For counting, a 3 in.

x 3 in. Nal(Tl) scintillation detector and a 400-channel pulse

height analyzer were used.

Due to the high manganese and low vanadium content in five

of the samples, interference was expected in the vanadium

40
analysis from the Compton continuum from the 0.845 MeV and

1.8l MeV gamma rays of manganese-56 . Therefore, several preliminary experiments were performed which showed that better

accuracy could be obtained when a complement subtraction technique was used. The following data shows the same sample

analyzed with and without subtraction.

Table 13. Comparison of Vanadium Results in SRM 1261

With and Without Complement Subtraction of the

Manganese Activity

Vanadium Concentration {%)

Sample No. Without Subtraction With Subtraction

1 0.0127 0.0113

2 0.0125 0.0108

3 0.0123 0.0107

4 0.0130 0.0113

5 0.0132 0.0106

6 0.0127

Average % V 0. 0127± . 0003a

. 0109± . 0004a

a

± values are ts//n g 95$ C. L.

The results obtained using the subtraction technique agree

with those from other analytical methods.

a. Experimental

Samples were analyzed as received. Approximately 100 mg

each of steel was sealed in a polyethylene snap-cap vial.

Solution standards of vanadium ( 99 • 9+% ) were pipetted onto

filter paper (free of vanadium) inside the vials and evaporated

The vials were then heat sealed. Copper foils were attached

to both the samples and standards.

The procedure used was as follows:

(1) The sample or standard was irradiated 1.0 minute.

(2) The sample was allowed to decay for 2.0 minutes.

During this time the flux monitor was removed and the

sample positioned on the detector.

41
(3) At the end of the decay time, the sample was counted

for 4.0 minutes clock time in the add mode. The live

time data is read out in channel zero.

(4) The vanadium is allowed to decay and manganese is

subtracted by accumulating for approximately 4.4

minutes in the subtract mode.

Manganese-56 (2.56 hours half-life) was counted after the V

and other short-lived activities had decayed. The peak areas

of both manganese and vanadium were integrated using Covell's

method [13]. The concentration of Mn for each sample was

then calculated using the standard-comparator method. Flux

variations from sample to sample were normalized using the

flux monitors. Self-shielding and gamma-ray attenuation was

calculated and found to be negligible for these conditions.

b . Results and Discussion

The analysis results for the six SRM's can be found in

table 16. Replicate samples agree very well with each other

and there appears to be no evidence of inhomogeneity within

the limits of the precision of the analysis.

Table 14. Results of Manganese and Vanadium Analysis

of SRMs 1261, 1265, 1155, l60b , 20g, 36b

SRM No . of Average %

Sample Samples Mn Found

1261 6 0. 63721.0027

1265 8 0.0074±.0002

1155 6 1.620 ±.034

160b 6 1.634 ±.030

20G 4 0.825 ±.051

36b 3 0.402 ±.014

No. or Average %

Samples V Found

5 0.0109 ±.0009 a

9 0.00054±. 00002

6 0.0482 ±.0023

6 0.04864±.0017

4 0.0013 ±.0004

3 0.0031 ±.0003

aErrors are reported as ts//n at the 95% confidence level.

(T. E. Gills and D. A. Becker)

42
3 . The Analysis for Lanthanum, Tantalum and Selenium

in SRM Special Iron No. 1265

The Office of Standard Reference Materials requested

analyses for La, Ta, and Se as part of acceptance testing of

a new SRM Iron. This particular iron (#1265) was the starting

material for a series of special irons (#1261-1265) and was

expected to have low concentrations of most of the elements

of interest.

a. Experimental

The samples, which consisted of iron turnings, were

weighed and encapsulated in polyethylene snap-cap vials.

Standards of the elements of interest were also prepared.

Copper foil flux monitors were attached to both samples and

standards for later normalization of the neutron flux values.

The samples and standards were irradiated for 4 hours in the

n 12 NRL Reactor at a thermal neutron flux of approximately oxlO

-2 -1

n.cm . s . After irradiation, the samples were brought back

to NBS and allowed to cool for about one day. A 1000 minute

count was made using a large volume Ge(Li) detector and a 409 6 channel pulse height analyzer. The resulting spectrum

was used for the lanthanum and tantalum analysis. As no

gamma-ray peaks for tantalum-182 were observed, an upper

limit for the tantalum content was calculated.

Due to the relatively low energy gamma rays and low

sensitivity for selenium-75, a chemical separation was necessary to determine selenium. The procedure used was the precipitation of selenium metal with hydroxylamine-hydrochloride

.

The samples were dissolved in a mixture of nitric and perchloric acids, with ^30 mg of selenium carrier present.

After complete dissolution, the solution was taken to fumes

of perchloric acid, making certain that nitric acid was present

until all carbon (or organic material) was oxidized off. The

resulting solution was taken up in 3M hydrochloric acid, and

^10 ml of a 25$ solution of hydroxylamine-hydrochloride added.

43
Several hours heating at 90°C assured complete precipitation

of the selenium. The solution was cooled, filtered onto a

glass-fiber filter pad, and counted. Tracer studies on

similar samples which were not irradiated showed the chemical

yield to be greater than 99%.

b . Results and Discussion

The lanthanum concentration found was 0.008 ppm. Since

lanthanum is one of the more abundant rare earths, this value

could be used to estimate the content of the other rare earth

elements

.

No tantalum was detected in these samples. The amount

which would have been seen under these conditions was calculated to be 0.2 ppm; therefore, the amount of tantalum present

is <0.2 ppm.

No selenium-75 was detected. The calculated upper limit

was found to be 0.3 ppm.

(D. A. Becker)

4 . The Determination of Homogeneity of Picogram

Quantities of Sodium by Neutron Activation Analysis

The NBS Standard Reference Material program has just

recently made available its first microstandard particles.

These microstandards are small (5-25y diameter) calcium

loaded ion exchange beads, and are certified for calcium

content [14]. The production and coordination for this SRM

was provided by the Separation and Purification Section of

the Analytical Chemistry Division.

The Activation Analysis Section was asked to determine

the homogeneity of sulfonation of these ion exchange beads,

since it was the only analytical competence with the capability of analysis at the picogram level for a convenient ion

with the requisite precision (2-5JO. The ion exchange- beads

were loaded with sodium for this analysis, since sodium can

be determined at the required levels while calcium cannot.

44
The homogeneity of the sodium (in units of picogram/u ) was

used to establish the uniformity of sulfonation of the ionexchange copolymer. Then saturation loading of the beads

with calcium ions yields a known amount of calcium per unit

volume of bead.

a. Experimental

The samples, 1-2 mg of bulk sodium loaded ion-exchange

beads, were encapsulated in quartz and irradiated for 5-7

hours in the glory-tube facilities of NRL Nuclear Reactor at

12 -2 —1

a thermal neutron flux of ^8x10 n.cm

_ .s . The samples

were immediately returned to NBS and H. Dixon, of the

Separation and Purification Section, selected individual

beads for analysis using micromanipulators. Each bead selected

was photographed under a microscope at a known magnification

and prepared according to the specified procedure [1*0 . Although

some irregularities in bead shape were noted, most beads were

spherical. The beads to be analyzed by activation analysis

were selected to be as nearly spherical as possible, since the

measured diameter was used to calculate the volume of the

resin bead.

After selection the beads were prepared for counting

by sealing the individual resin beads between two 10 mil thick

plastic scintillator disks. Thesp fiisks were than mounted on

the 3/4" phototube assembly and counted using the 3yy sum

coincidence spectrometer described elsewhere in this report

(see section 3.B.1).

Three of the samples were positioned between two 4"x8"

Nal(Tl) gamma-ray scintillation detectors. The system logic

requires that a beta particle and two gamma-rays (one in each

detector) be detected within the resolving time of the coincidence unit. When such an event occurs, the pulses from the

two gamma-ray detectors are summed and presented to the input of a pulse height analyzer. The pulses from each of the

three samples are stored in a separate quadrant of the analyzer. 45
Since the sum of the two gamma rays of sodium-24 is 4.12

MeV, the energy range of the pulse height analyzer was set

at 0.1-4.5 MeV. Most samples were counted for 900 minutes,

and at least two sets of three samples were counted from

each irradiation. A total of 42 resin beads, from five

different irradiations, were analyzed for their relative

sodium content. The standards used were several small

granules of sodium sulfate (53-132 ug Na) , and were irradiated

for 1 minute in the same irradiation facility. All samples

and standards had a copper-lead alloy flux monitor attached

to normalize irradiation conditions. After irradiation,

individual granules of the sodium sulfate were sealed in the

plastic scintillators and counted exactly the same as the

samples. Considerable decay was allowed for the standards

so that the count rate was comparable to the samples, since

serious errors can result when a comparison is made between

sources with large differences in counting rate.

For the analysis, the total counts from 0.1-4.5 MeV

were used, with an average background subtracted. The total

background from 0.1 MeV-4.5 MeV averaged 3 counts per hour.

The decay of several of the samples was followed for several

half-lives and the purity of the sua±um-24 radioactivity verified. In addition, a gross sample of the irradiated resin

beads were counted on a 45cc Ge(Li) semiconductor detector

and 4096 pulse height analyzer. No radioactivity other than

sodium-24 was observed.

b . Results and Discussion

The results of the analysis are shown in table 15. The

sodium concentration found (the ratio of sodium to unit volume) was approximately constant, averaging 0.136 pg/y with

an uncertainty of 2% (ts//n @ 95$ confidence level) . The

size distribution of samples analyzed is found in figure 20.

Considering the difficulty in locating and separating the

individual beads, the size distribution seems to adequately

46
o

UJ

IM4

Z

<o

<

CD 2

u.° I d vMmvWfflm liLI

fcaa Kfiifa llLiIgkj&a

10 12 14 16 18 20

BEAD DIAMETER (//.)

22 24 26 28 30

Figure 20. Size distribution of microstandard beads analyzed

represent the bead size range in the bulk sample. There is

no immediately apparent systematic bias due to sample size.

However, it can be seen from table 16 that the precision of

the sodium concentration decreases significantly with bead

diameter. This is to be expected, since the counting statistics are greater (fewer total counts) and the uncertainty in

the measurement of the bead diameter is greater. It is not

presently known whether some of the decrease in precision of

the smaller beads is also due to an inhomogeneity in sulfonation in the resin bead itself.

The precision of the technqiue is estimated at ±2%, as

determined by repetative counting of sodium-24 and cobalt-60

standards. The accuracy of the analysis is estimated to be

±10?, due primarily to the possibility of error introduced

in the vastly different irradiation times for standard and

samples. Additional work will be done on some larger beads

and on an accurate determination of absolute sodium sensitivity

One possible bias which has not yet been investigated is

the effect of irradiation on the resin bead size. Since

the size measurement for these analyses was made after the

irradiation, the radiation effects on the bead, such as

increased cross-linking of the polymer produced by the large

neutron and gamma-ray fields in a nuclear reactor, could have

a significant effect on the final bead size. This effect will

be evaluated in the near future.

47
Table 15. Sodium

Bead Run Net

No. No. Counts

1376 1-1A Ul,260

1380 1-1B 9,315

1378 1-1C 7,099

1379 1-2B 9,367

1394 2-1A 10,128

1395 2-1B 5,085

1396 2-1C 12,432

1391 2-2A 7,806

1392 2-2B 9,703

1397 2-2C 6,231

1408 2-3A 25,996

1409 2-3B 33,Q32

1411 3-1A 2,271

1412 3-1B 501

1414 3-1C 7,348

1417 3-2A 1,185

1418 3-2B 1,370

1419 3-2C 2,516

1420 3-3A 1,461

1421 2-2B 5,000

1422 3-3C 5,273

1428 4-1A 1,375

1427 4-1B 325

1424 4-1C 3,541

1426 4-2A 860

1437 4-2B 3,285

1432 4-2C 2,364

1433 4-3A 4,207

1434 4-3B 13,847

1436 4-3C 2,197

1469 5-1A 61,653

1472 5-1B 61,771

1465 5-1C 93,374

1475 5-2A 27,095

1474 5-2B 25,336

1471 5-2C 34,495

1477 5-3A 25,358

1473 5-3B 37,188

1481 5-3C 33,475

1480 5-4A 12,104

1479 5-4B 22,053

1470 5-4C 25,697

Sodium Content of Microstandard Beads

Sodium Sodium

Diai.. ^ter oer bead Cone . (y) (pg.) pg./(v

16.776 320 0.129

9.793 66.3 0.135

9.327 53.6 0.126

12.659 148.1 0.140

10.877 86.9 0.129

8.247 39.9 0.136

11.549 103-6 0.128

12.950 152.3 0.134

13.770 173-0 0.127

12.132 118.1 0.126

26.315 1354 0.142

28.200 1617 0.138

6.470 20.2 0.143

4.012 04.08 0.121

9.553 63.6 0.139

6.691 22 . 9 0.146

6.510 24.2 0.167

8.646 47.2 0.139

9.701 61.6 0.129

13.786 192.9 0.141

14.318 216.4 0.141

5.896 14.7 0.137

3.755 03.19 0.115

7.925 36.9 0.142

6.356 20.2 0.150

10.201 70.5 0.127

9.229 53-9 0.131

15.253 251.7 0.136

22.088 757.8 0.134

12,555 127.9 0.124

22.730 861 0.140

22.107 789 0.139

25.896 1269 0.140

21.688 745 0.140

20.787 637 0.136

23.506 923 0.136

19.204 518 0.140

21,363 695 0.136

20.952 666 0.138

21.880 754 0.138

25.746 1256 0.141

27.749 1557 0.139

48
Table 16. Statistical Variation of Sodium Content

With Microstandard Size

Average Na Standard ts//n

Microstandard No. of Content Deviation at 95%

Size Diameter Beads (pg/y 3

) at 95% C.L. C.L.

All Beads

(3.7y to 28. 2y) 42 0.136 13%

Large Beads

(20. Oy to 28. 2y 14 0.138 3.2% 0.9%

42 0.136

14 0.138

13 0.132

5

15 0.137

Medium beads

(10. Oy to 19. 9y) : 0.132,, 4.9% 3-5%

Small beads

(3-7y to 9.9v) 15 0.137 18% 5.1%

(D. A. Becker)

5 . The Determination of Impurities in Clinical Standards

The Office of Standard Reference Materials has begun a program of using a number of standards for clinical chemistry

The first of these standards, urea, createnine, uric acid,

and calcium carbonate, have been issued. Because of the

importance of trace metal content as possible enzyme poisons,

the Activation Analysis Section was asked to survey these

standards for trace metal and also for chlorine

a. Procedure

Samples of 1.0 gram of each material were irradiated for

12 —2 —1 1.0 hours at a flux of approximately 8x10 n cm s in the

in-core tube of the Naval Research Laboratory Reactor. Standards of the elements listed in tables 17 and 18 were irradiated simultaneously. Copper foils were attached to both

samples and standards for flux normalization. For the determination of chlorine and aluminum, shorter irradiation and

decay times were used.

49
Table 17. Analytical Conditions for the Determination

of CI, Al, and Na in Clinical Standards

Peak

Energy

Irradiation Decay Counting Used Counting

Element Isotope Time Time Time (MeV) System

p o Aluminum Al

Chlorine 3 CI

Sodium 24Na

1.0 min. 1.0 min. 4.0 min. 1.78 Nal(Tl)

Clock

10 min. 10 min. 4 . 0. min

.

Clock

1.6 Nal(Tl)

1.0 hr. 1.0 min. 100 min. 1.36 Li(Ge)

Live

Table 18 . Impurity Content of Clinical SRM f s Concentration of Impurities v Cppm)

Calcium

Element Urea Createnine Uric Acid Carbonate

Sodium 9 30 170 0.5

Copper .27 0.20 2.4 0.9

Zinc .24 < .5 < .2

Manganese .008 .07 .28 0.6

Gold .0001 - - - Arsenic < .01 < .01 < .01 - Antimony < .01 < .01 < .01 - Chlorine <1 330 - < .01

Aluminum .9 - - Potassium — 11

50

(T. E. Gills, D. A. Becker)
b

.

Sample Dissolution and Standard Preparation

After irradiation, each of the materials was allowed to

decay for 2.5 hours. The samples were then dissolved in a mixture of HNO~ and HCICk to destroy any organic material.

The samples were then evaporated to incipient dryness and

made to 100 ml volume with deionized water. Ten milliliter

aliquots were taken for analysis

.

The standards were dissolved and made to 100 ml volume

with deionized water. Chlorine and aluminum were determined

nondestructively

.

c

.

Separations

Gold, Copper : To three 10 milliliter aliquots of each of

the samples 25 ml of a 2M solution of hydrochloric acid was

added. Ten milligrams each of copper and gold carrier was

added to each of the samples. Ten milliliters of 50% thioacetamide was then added to precipitate the sulfides. The

samples were heated to coagulate the precipitate and allowed

to cool. The samples were then filtered, using glass fiber

filter pads, dried and counted.

Manganese : Another three 10 ml aliquots were made to

25 ml and 6N HNO.,. Ten mg of manganese carrier and approximately 25mg of potassium bromate were added. The samples

were heated until the manganese dioxide coagulated and then

filtered onto a glass fiber filter pad, dried and counted.

Zinc : The remaining solution was divided into 10 ml aliquots and zinc carrier (20 mg) was added. The solutions were

then- made to . 01N in sulfuric acid. Ammonium phosphate was

then added and the zinc precipitated as zinc ammonium phosphate. The precipitate was filtered onto glass fiber filter

pads, dried and counted.

The elements and standards were counted using a 3 in. x

3 in. Nal(Tl) detector and a ^00-channel pulse height analyzer

Half-life determinations were made for verification of spectral purity.

51
The determination of chlorine, aluminum and sodium was

made by varying the irradiation and decay times. Table 19

shows the procedure user)

.

Table 19. Results of Gold Analysis in the SRM

Cobalt-Aluminum Flux Wires

Sample yg gold/gm wire (ppm )

3 4.12

8 4.29

11 4.26

Average 4.2±0.3 a

Average is ts//n at 95$ confidence level.

Standards of each of the above elements were irradiated

and counted at the same geometries and time sequences. Flux

monitors were used to normalize flux conditions . Live time

used for aluminum and chlorine analysis were those in channel

zero of the analyzer.

d. Results

The concentration of each element was determined using

the standard-comparator method. The results of the analysis

are given in table 18. Upper limits were calculated for

arsenic and antimony.

6. The Determination of Gold in the SRM Cobalt - Aluminum Flux Wires

During the cobalt analysis of these SRM neutron flux wires

a significant gold impurity was discovered. The gold radioactivity produced with short irradiation and decay times was considerably greater than that due to cobalt-60. Since the gold

was present in the wire, the homogeneity of the gold was investigated to see if it could be used as a second flux monitor

element

.

52
a. Experimental

Three samples of cobalt-aluminum flux wire (^4 mg each)

were cut, encapsulated and irradiated for 5 minutes in the

NBSR pneumatic tube facility, along with a gold standard.

24 After irradiation the Na produced by the fast neutron

27 24 reaction Al(n,a) Na was allowed to decay for several days.

At the end of this time, the samples and standards were

counted. The 0.412 MeV gamma-ray of the samples was followed

for several weeks and the decay curve of the integrated peak

plotted to verify the identification of and gamma-ray purity

198 of the Au. The experimental half-life of 65. h agreed

well with the literature value of 64.8 h [15].

Special mention may be made here about the danger of

using gold metal (foil or wire) as a standard for neutron

activation analysis. Our results show a significant (19%)

neutron self-shielding effect for 1 mil thick gold foil.

This agrees quite closely to the 23% self-shielding for 1

ml gold foil as found by Reynolds and Mullins [16], Of course,

the exact self-shielding effect depends upon the neutron

energy spectrum in the irradiation facility used. In addition,

Reynolds and Mullins found a significant (^5%) thermal neutron

enhancement effect for aqueous samples ca. 1.5 ml [16], Thus,

it has been our practice to prepare gold standards by pipetting

a small aliquot (50X) of a freshly prepared gold solution

onto a "bed" of filter paper (^2mm thick) inside a polyethylene vial, allowing the water to evaporate, and then covering

with a fresh disk of filter paper before sealing the vial.

This extra filter paper assures the gold will not leave

the "bed" area, and thus maintains a sharply defined counting

geometry. The wire samples were coiled to produce a similarly shaped counting source. Blank samples of the vial and

filter paper showed no detectable gold radioactivity for the

conditions used for this analysis.

53
The gold homogeneity data was obtained from the same

data accumulated for the cobalt homogeneity analysis. Four

different sets of data were used, and the results are shown

below.

b . Results and Discussion

The results for the gold analysis are found in table 19.

Three different cuts were analyzed, and averaged 4.2±0.3 ppm

gold. This value agreed well with an analysis by spectrophotometry, which gave a value of 4.5 ppm gold.

The results of the homogeneity study are found in table 20.

The values given are in terms of relative gold concentration,

since no gold standard was irradiated with these samples.

This second set of results show good homogeneity, especially

for this level of impurity, giving an uncertainty of 1.3$ for

ts//n at the 95% confidence level.

Table 20. Homogeneity of Gold in Cobalt-Aluminum Flux

Wires

Average

Relative Gold

Concentration

0.993

1.013

Bag No. Sample No.

Relative Gold

Concentration

11

11

11

11-1

11-2

11-3

1.012

0.961

1.005

12

12

12

12-1

12-2

12-3

1.032

1.007

1.000

13

13

13

13-1

13-2

13-2

0.993

1.036

0.992

14

14

14

14-1

14-2

14-3

0.991

0.976

0.996

1.007

0.998

Average of four different bags 1 . 0008±0 . 0133a

Uncertainty values of ts//n at 95% confidence level.

54
7. Analysis of Simulated Lunar Glasses

a. Introduction

The Office of Standard Reference Materials has begun a program to make available samples of glass to which traces

of 61 elements have been added. As part of this program,

studies are being carried out to characterize the base glass

before the addition of these trace elements. The base glass

is 12% SiO pJ 1H% Na

p O, 12$ CaO, and 2% Al p O . When this glass

is irradiated for sufficiently long times to permit determination of impurities at the ppm or ppb level, 20 ior more mCi of

24Na are formed. Recently Girardi and Sabbioni [12] have

described the use of hydrated antimony pent oxide (HAP) for

selective removal of sodium from solutions containing many

24 other elements. We have used this HAP for removal of Na

from the irradiated glass samples.

b

.

Experimental

The samples are dissolved in HF-HClOj, after rinsing in

alcohol and dilute HN0_ to remove surface contamination. The

dissolution is somewhat lengthy because of formation of insoluble CaP

p

which precipitates onto the glass surface and

inhibits further attack of the SiO

p by HP. By alternately

adding small portions of HP and evaporating to fumes of

HClOj. to decompose CaF p

, dissolution of a 1 gm disk of glass

^ 1 cm in diameter can be accomplished in 1-2 hours. The

solution is then taken to fumes of HC10|. and fuming continued

for at least 30 minutes to remove SiFj. and all traces of HF

and to reduce the volume of HClOj,. The cooled residue is

taken up in 15-20 ml of 12 N HC1. The solution is passed

through a HAP column containing ^ 5 gm HAP from which the

fine particles have been removed by slurrying with 12 N HC1.

The column is washed with a small amount of 12 N HC1 and

the effluent is collected in a polyethylene bottle and counted

directly with a Ge(Li) detector. The decontamination observed

is such that the HAP column may be reading as much as 1-2 r/hr

55
24 while no Na peak can be detected in the effluent after counting for 100 minutes. Occasionally, poorer decontamination

2 3x 29 (factors of only 10 -10 ) from Na is obtained and it may be

necessary to pass the effluent through a second HAP column.

The consistency of the results obtained for the various elements

present (except for K) regardless of the number of HAP columns

used confirms Girardi's results on the lack of retention of

elements other than Na and Ta by HAP. A typical spectrum of

the Na-free glass solution is shown in figure 21. Peaks corre72 p l40 T 42 T^ 47, spondmg to Ga, La, K, Ca (from the CaO in the matrix)

1?? 1?4 64 198 47 47 Sb - Sb, Cu, xy Au, and 'Sc (from decay of 'Ca) can

be readily observed. Table 21 summarizes the results obtained

to date. These results were all obtained using our new 47-cc

Figure 21. Gamma-ray spectrum of sodium- free glass solution

5 e
Table 2 1. Results of Analysis of Ease Glass, ppm

Sample Ca Au Sb La Ga K

#61 0.91 0.22 0.064 0.032 0.34 25

#107 1.09 0.25 Not Determined

#107A-1 0.95 0.26 0.076 0.033 0.33 8 a

#107A-2 0.96 26 0.69 0.034 0.36 28

#6lA 1.06 0.20 0.067 0.023 0.32 24 c

#6lA-2 1.29b 0.20 0.077 0.034 0.33 33

#87A-1 1.14 0.21 - 0.033 0.36 38

#87A-2 1.06 0.17 0.074 0.028 0.34 24

#31A-1 1.03 0.22 0.075 0.030 0.32 22

#31A-2 1.01 0.18 0.056 0.024 0.32 20 C

#131A-1 1.05 0.18 0.064 0.030 0.34 29°

#131A-2 0.92 0.23 0.044 0.026 0.27 8

C

#8A-1 1.04 0.18 0.064 0.026 0.32 8

C

#8A-2 0.99 0.20 0.066 0.027 0.32 - #4la-l 1.06 0.26 O.O65 0.028 0.35 26

#4lA-2 1.15 0.24 0.064 0.030 0.36 34

#117A-1 1.09 0.19 0.053 0.026 0.29 22

#117A-2 0.08 0.24 0.05] 0.030 0.32 32

a

3 HAP Columns

Some ~ Na present, possibly causing high results.

C

2 HAP Columns.

Ge(Li) detector, whose resolution is shown by figure 22 which

is a plot of the 0.835 MeV peak of 72 Ga and the 0.847 MeV peak

56 of Mn as observed in this glass at an earlier time than represented by figure 21. A paper describing the results on the

base glass has been submitted for presentation at the New York

ACS meeting and for publication in the literature. The work

will be continued incorporating the doped glasses

57
UJ>

<

UJ

1800 1 1 I 1 1

\ 0.835 MeV

1

72ca

1600 ""

1400 :

< 0.847 MeV

• 56 Mn -

i 1200 " —

1000 =

' -

800

i -

600 :

-

400 : J

\l

-

200 «~»v> M/s -

i i i i i

1 630 640 650 660

CHANNEL NUMBER

670 680

Figure 22. Resolution of gallium-72 and manganese 56 using

a Ge(Li) detector

(B. A. Thompson)

8. Determination of Tungsten in SRM 480

Tungsten was determined by neutron activation analysis in

the small samples of W-Mo alloy described above. The analysis

was carried out twice. The first measurements were made

on aliquots of the dissolved samples which had been irradiated in the glory tube facility of the NRL Reactor. The

aliquots were counted using a 3 in. x 3 in. Nal(Tl) detector

58
and a 256 channel pulse height analyzer and the area under

the 0.480 MeV and 0.686 MeV peaks used as a measure of the W

present

The results obtained were:

Table 22. Results of W Analysis of W-Mo Alloy

% of W

1 69.4 ±2.3a

2 73.1 ±1.8

3 68.4 ±0.9

4 67-75±1.2

Limits quoted are ts/n for the 95% confidence level;

n = 6.

These results, added to the Mo results reported above,

gave totals of less than 100$. It was noted during dissolution that small black flakes were present which were insoluble in HF-HNO- but which disappeared after fuming with HClOj,,

indicating possibility of something other than W and Mo being

present. However, it was decided to repeat the analysis as a

check. Five additional samples were irradiated in the North

rabbit of the NRL Reactor where the thermal-to-fast ratio is

much higher and W was determined nondestructively . By using

these two different irradiation facilities the magnitude of

any resonance self-shielding could be observed.

The results obtained were:

68.92±1. 265S W

confirming the previous values and showing that the resonance

contribution is negligible. Results obtained by differential

spectrophotometry also showed the sum of the W and Mo concentrations to be <100$ in these samples.

(B. A. Thompson)

59
E . Service Analyses

1. Determination of Chlorine in Al o 0„ and Ta Cv by 2_3 2 5

J Activation Analysis

Several samples of single crystals of A1 ? and of vaporgrown Ta^Oj. were analyzed for chlorine by nondestructive

neutron activation analysis. This determination was requested

by the Inorganic Materials Division.

The samples, weighing about 200 mg each, were irradiated

for 1 minute in the NRL rabbit. Two NH^Cl standards were

irradiated simultaneously with the samples. Both samples and

standards were contained in polyethylene snap-cap vials with

a Cu flux monitor taped to the outside of each vial. The

samples were allowed to cool for 30 minutes, after which, they

were transferred to new snap-cap vials and counted with a 3 in.

x 3 in. Nal(Tl) detector and a 400 channel pulse height anao o lyzer. The 1.60 and 2.17 MeV y-rays of 3 CI were used as a measure of the CI present. The results showed less than

the Ta^Cv.

2 5

(B. A. Thompson)

0.01$ CI in the Al o o and 0.02-0.03$ CI in the Ta^Cv

.

^3 2 5

2 . Determination of Selenium in Electrodeposited Copper

The Activation Analysis Section was asked to determine

selenium at the ppm level in a matrix of electrodeposited

copper for the Electrolysis and Metal Deposition Section of

the Metallurgy Division at NBS . There was a question as to

whether the selenium present in the plating solution was

finding its way into the copper deposit.

There are three distinctly different ways of performing

this analysis, the choice being dictated by the sample characteristics. These are nondestructive analysis using selenium77m (t., ,p=17s ) , nondestructive analysis using selenium-75 (t-t/?

=120d) , and destructive analysis with chemical separation of

selenium-75

•

60
a. Experimental

The samples received were six pieces of copper metal, each

several square inches in area and approximately .030 in. thick.

The samples were cut in half to obtain duplicates for analysis.

They had to be folded once to fit into the irradiation facility.

A test was made on one of the samples to see if selenium77m could be used for the analysis . The large amount of

copper radioactivity produced from even a very short irradiation (^5 sec.) completely obscured any selenium present.

Therefore, selenium-75 was used for the analysis.

The samples and a selenium standard were encapsulated in

pre-cleaned polyethylene vials, and irradiated for two hours

n 12 -2 -1

at a thermal neutron flux of ^oxlO n.cm s . After irradiation, the samples were set aside for approximately one month

to permit the copper radioactivities to decay. The selenium

standard and copper samples were then counted using a large

volume Ge(Li) y-ray detector. The six selenium y-rays at

the proper energies and in the correct proportions (97keV, 3-3%;

121keV, 17$; 136keV, 57$; 265keV, 60%; 280keV, 25%; and 104

keV, 12%) served as positive verification of the selenium

content. The 265keV. peak was chosen for integration because

of its higher abundance. Since the gamma-ray energies are

quite low and the copper samples quite massive, it seemed

likely that some gamma ray self-absorption was occurring in

the sample. Calculations of estimated self-shielding showed

that for a copper sample 1 mm thick the error would be ^10%.

Also, the counting geometry of the sample and standard were

much different due to their different sizes, and this would

be expected to create an additional bias. To determine a

correction factor for these effects, two of the samples were

carried through a chemical separation after the Ge(Li)

counting.

The separation used was reduction of the selenium to the

metal. After dissolution of the samples in dilute nitric

61
acid with added selenium carrier present, several ml of perchloric acid was added and the samples evaporated to fumes of

perchloric acid. The samples were taken up in ^10 ml of dilute

HC1, and an excess of 25% hydroxylamine hydrochloride solution

added. The solutions were heated at ^90°C for 1-2 hours,

filtered through glass fiber filter pad, washed and mounted

for counting. The selenium standard was treated identically.

The separated samples and standards were counted on the Ge(Li)

detector, this time with the same sample-detector geometry.

Decay corrections were made in all cases

.

b . Results

The results of this analysis are shown in table 23. These

values include the correction factor determined by the chemical

separations. The chemical separation has a yield of 99 • 9+^ as

determined by trials with synthetic samples . This agrees with

the work of Gorsuch[17], who found that selenium does not

volatize when heated in mixtures of nitric and perchloric

acids. The estimated accuracy of the analysis is ±10%.

Table 23. Results of Analysis for Selenium

Concentration Selenium

Sample No. Weight (g) (ppm)a

301-1 A 0.6319 10.3

303-1 B 0.6168 11.8

303-2 A 0.6234 11.0

303-2 B 0.6236 11.2

303-3 A 0.6771 11.8

303-3 B 0.6448 10.6

172-1 A 0.5883 0.8

172-1 B 0.5315 0.8

172-2 A 0.5790 0.4

172-2 B 0.5755 0.5

172-3 B 0.5986 0.7

(D. A. Becker)

3. Estimated accuracy of analysis is ±10%.

62
3 . The Determination of Selenium by Neutron Activation

Analysis

Selenium was determined in a number of organic and inorganic materials by nondestructive neutron activation analysis.

These analyses were performed for the National Center for Urban

and Industrial Health of the United States Public Health Service

a. Experimental

The samples were analyzed as received. One hundred mg of

each of the samples were encapsulated in a quartz ampoule.

Selenium metal (99.9+%) was also encapsulated in quartz to be

used as a standard. Copper foils were attached to both samples

and standards for flux normlaization.

Both samples and standards were irradiated for five hours

12 — 2 —1

at a flux of approximately 8+10 n.cm sec in the glory

tube of the Naval Research Laboratory Reactor. After irradiation the samples were removed from the quartz ampoules and

placed in 2/5 dram poly vials for counting. The counting was

done on a 30 cc Ge(Li) detector in conjunction with a 4096

pulse height analyzer. Verification of presence or absence

of selenium was easily made using the four prominent selenium75 gamma rays.

b

.

Results

The results of the analysis are given in the following table

For compounds where selenium was not detected upper limits are

given. In most cases greater sensitivity could be obtained by

radiochemical separations of the selenium before counting. The

precision and accuracy of the results are estimated to be ±-10

percent for the standard error.

63
Table 24. Selenium Content of USPHS Samples

Compound Amount Found (ppm )

B-naphthylamine < 2

l-amino-2-naphthol < 1

Ethyl Carbomate < 1

9, 10-dimethyl-l , 2-benzanthracene < 1

3-methylchelanthrene < 1

Beryl Ore < 1

Low Fired Beryllium Oxide Lot #840 < 1

Benzo (a) pyrene 116

Betox 6 < 1

Beryllium Sulfate < 1

Beryllium hydroxide < 1

Nickel Sulfate < 1

(Thomas E. Gills)

64
4. ACTIVATION ANALYSIS WITH THE COCKCROFT -WALTON

NEUTRON GENERATOR

A. Introduction

In the time period covered by this report, research into

possible application of the facility using 3 MeV neutrons

? 3

from the H(d,n) He reaction was emphasized. Information in

literature pertaining to the use of 3 MeV (D-D) neutrons is

virtually non-existent. Consequently, there was a need for

quantitative analytical data which could provide a base for

the evaluation of errors due to D-D neutron activation in

14-MeV neutron activation analysis, and to project possible

new applications for 3 MeV neutrons for analysis purposes.

As was reported previously [1,18] the analysis of mixed

halogens in photographic emulsions was made possible by the

combined use of 14- and 3-MeV neutrons. This particular

study provided sufficient incentive to initiate a systematic

investigation for the measurement of D-D neutron activation

response of most of the elements in the periodic table, with a

view to establishing a catalogue of spectral and sensitivity

data. Such catalogues for 14-MeV neutron activation analysis

are quite common, however, one for D-D neutrons has not appeared in literature. Excluding the noble gases and the

naturally radioactive elements, about seventy pure elements

have been irradiated and gamma-ray spectra measured under

normal experimental conditions, and sensitivities have been

calculated. These data are presently being put into a form

most useful to the analyst. Some typical information is given

in this report.

Before information obtained from the above survey can be

used in the assessment of systematic errors due to D-D activation in 14-MeV neutron activation analysis, the D-D flux

build up with beam time on a tritium target must be determined precisely. Preliminary work has begun and special

targets to be used in this study are presently being manufactured. These targets closely simulate a normal 0.8 Ci/cm

65
tritium target, the difference being the replacement of tritium

by an equivalent concentration of hydrogen. By irradiating

these hydrogen loaded targets, the build up of the 3 MeV

flux as a function of integrated beam dose can be measured.

These flux measurements can be used to predict D-D interferences in any analysis performed with 14-MeV (D-T) neutrons.

In addition, a research program has been initiated to

evaluate thermalization of D-D neutrons using either D„0 or

HpO as moderators, with a view to the utilization of the

thermal flux for the determination of key elements in geological samples. Although the incident D-D flux is about 200

times less than the D-T flux for our facility, thermalization

of D-T neutrons is considered to be a very inefficient process

due to significant contributions from the fast flux which can

present serious interference problems. An analysis using

thermalized D-D neutrons was very successful for the determination of sodium in a metallo-organic Standard Reference

Material (SRM), and results from this study should yield many

useful applications.

During the past year the neutron generator has been used

for the analysis of four organometallic oil additives. Experimental details and final results are reported below. Comparisons with results obtained using chemical techniques on the

same materials show good agreement. In addition to these

analyses, service analyses have been provided to other sections

of the Analytical Chemistry Division and some outside agencies.

Notable among these are, the determination of oxygen in styrenedivinylbenzene copolymers used for the preparation of micro

standard SRMs, and the analysis of oxygen in low-oxygen polyethylene vials manufactured by a special process.

Much of our total effort during the past year was concentrated upon the improvement and updating of the existing

facility, which was prompted by breakdown of key ancillary

components of the major facility, resulting in unavoidable

66
downtime. Although repair in most cases represented only

10$ of the total downtime, transportation of the defective

device to and from the manufacturer accounted for the major

portion of the delay. In order to avoid these delays in the

future, we have gradually upgraded the facility itself as

well as obtained sufficient knowledge to perform many of the

repair functions in the laboratory. It is hoped that the

entire facility will be self-sufficient in the near future.

Some of the problems experienced and of interest to accelerator users have been described.

In the interest of maximum safety for all pei-sonnel in

the vicinity of the neutron generator, additional 'fail-safe'

devices have been installed and a rigid procedure for tritium

contamination control established. In this respect the operational characteristics of the generator have been investigated

from which optimum operating information can be obtained.

Health physics survey information of the biological shield

has also been given together with results of tritium monitoring of key accelerator components exposed to this environment over a long term period.

(S. S. Nargolwalla)

B. Research Activities

1 . Experimental Spectra of the Elements and Calculated

Sensitivities for 3-MeV Neutron Activation

In laboratories, both industrial and research, using

neutron generators for activation analysis, effort has been

primarily focused on the use of 14-MeV neutrons produced from

the D-T reaction. The reason for this preferential use of

the small accelerator is generally attributed to the high

14-MeV neutron output as compared with neutron outputs from

the D-D reaction. Depending upon the acceleration voltage,

target thickness and experimental configuration of the irradiation assembly, the 3-MeV neutron output from the D-D reaction

is about 200-300 times lower. In certain applications, however,

67
3-MeV neutrons can be extremely useful. For example, undesirable radioisotopes induced only by 14-MeV neutrons due to

the high threshold energies of the nuclear reactions, cannot

be produced if D-D neutrons are used. The elimination of such

interferences permits accurate analyses for major constituents

which can be activated with 3 MeV neutrons. Further, certaj 1" elements have relatively large cross sections for (n,y) or

(n,n'y) reactions at 3-MeV when compared to those for the

matrix. If such samples were irradiated with thermal neutrons,

matrix activation would prohibit the analysis. This "signalto-noise" enhancement with 3-MeV neutrons can be exploited

for certain applications. As described previously [1,18] the

rapid and accurate determination of mixed halogens in a photographic emulsion matrix was possible only by the combined

use of 14- and 3-MeV neutrons, because of the analytical constraints viz 3 rapid and nondestructive analysis. Although

such applications may be limited in number, the importance of

this aspect of accelerator usage was recognized and, we' believe, justifies a detailed study. To provide solid groundwork for this study, a systematic investigation of D-D activation of the elements is necessary. This information can

then be used to predict possible systematic errors in D-T

activation from products induced by D-D neutrons from the D-D

neutron flux build up as a function of deuteron beam time on

a tritium target . This type of interference has been recognized for a long time, however, published data are confusing

because of the lack of definitive information about target

loading, beam characteristics and sample assemblies. All of

these factors influence the flux build up from the interaction

of a deuteron beam with embedded deuterons in a tritium target.

A knowledge of 3-MeV neutron activation spectra of the

elements will be useful for evaluating this interference and

may open up useful applications of 3-MeV neutrons. The calculation of experimental sensitivities under average laboratory

68
conditions would be useful for the assessment of D-D interferences in D-T analyses. Much of this information has been

published for thermal and 14-MeV neutron irradiations, but

little or no information for 3-MeV neutrons has yet appeared.

Since reaction cross sections for 3-MeV neutrons are virtually

non-existent in the literature, we are systematically examining the gamma-ray spectra of each element under conditions

normally found in most neutron generator facilities.

The investigation reported here involved the measurement

of the gamma-ray spectra of seventy odd elements, excluding

only those elements which are naturally radioactive and the

noble gases. Prom the spectral data an experimental sensitivity was computed. The results from the entire study are

presently being compiled for catalogue-type presentation.

A typical gamma-ray spectrum is shown and useful analytical

data are given below.

a

.

Experimental Facility

The neutron generator facility, which includes the neutron

generator, pneumatic-tube transfer system, sample irradiation

assembly, sequential programmer and associated counting and

readout apparatus have been discussed in a previous report

[1] . A fast point plotter was used to produce the spectra

2 and deuterium targets (0.6 cc per cm ) were utilized for the

production of 3-MeV neutrons.

b

.

Procedure

The element or a pure compound of the element was irradiated in a D-D flux. The maximum limit placed on irradiation

time was saturation or 20 minutes, whichever was shorter.

In general, counting was started usually 0.5 min . after

termination of irradiation and the counting time was the

same as the irradiation time. Energy scale used for plotting

varied from 2.5 keV per channel to 20 keV per channel.

Printed teletype outputs were used to compute the experimental

sensitivities. This method involved integration of pertinent

69
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70
peaks by a simple base line subtraction technique to obtain

the net counts, and comparing the sensitivity per gram of

6 -2 -1

pure element normalized to a D-D flux of 10 n.cm . s . The

observed sensitivity was corrected for sample attenuation

using a method previously described [19] . Depending upon the

amount of material available, three different types of samples

were used Those listed under code A (see table 25) were packed

in 8 cc capacity polyethylene vials. For code B type samples,

1.6 cc vials were used. Partially filled 1.6 cc vials are

listed under code C. The absolute neutron flux was determined

so that subsequent irradiations could be interrelated by using

counts as registered by a neutron monitor. The stability of

the neutron monitor was checked daily by irradiating a sample

of ammonium iodide and monitoring the (n,y) activity from the

-| pQ 1 pO

25 minute ~ I activity under standard conditions. The ~ I

activity was related to the neutron monitor count. Throughout these experiments the neutron monitor exhibited good

stability

.

c . Determination of Absolute D-D Neutron Flux

For this determination, a knowledge of the gamma-ray

counting efficiency of the twin-detector system described

previously [1] was necessary. This problem was quite complex because of the source-detector geometry shown in figure

54

23. A cylindrical standard source of Mn closely approximating the sample in geometry was counted. The 0.84 MeV

Figure 23. Source-detector

counting geometry

71

SAMPLE

/CAPSULE

LUCITE

BETA

ABSORBER

RETURN GAS
gamma-ray peak was integrated and the detector efficiency

calculated was 7

.

o% . Since the sample thickness can introduce systematic errors, this efficiency could only be related

to a given sample by making appropriate corrections . A

sample of Al

p

O^ was irradiated and the . 8 ^4 MeV gamma-ray

3

27, peak resulting from the 9-5 minute activity ~

' Mg was counted.

After peak integration, the counts were corrected for both

incident neutron and counted gamma-ray attenuation using the

appropriate attenuation correction factors D-91. Using the

27 27 published cross section value of 1.5 mb for the Al(n,p) Mg

reaction, the absolute flux was calculated and related to

the neutron monitor counts recorded during the same irradiation. The neutron flux was also determined using the

Y(n,n'y) Y reaction (a=150 mb) . The agreement between

the flux values obtained by both methods was within 5% of the

mean. The location of the neutron monitor with respect to the

irradiation geometry is shown in figure 24. The ratio of the

calculated flux to the counts per minute registered by the

neutron monitor was 60. This factor was used for the computation of experimental sensitivities for all elements.

A typical example of the presentation of the data follows:

In figure 25 the gamma spectrum of barium is shown. Useful

nuclear analytical data related to this gamma spectrum is

BIOLOGICAL SHIELD /NEUTRON

MONITOR r1 NEUTRON

, I GENERATOR

Figure 24. Irradiation geometry showing location of neutron

monitor

72
10 —

uJ l0 IX

crUJ

0.

CO 10'

o

10

j H

~" A

./ \

\ .•

BARIUM

(ENERGY SCALE: 5keV

PER CHANNEL)

• v.

40 80 120 160 200

CHANNEL NUMBER

Figure 25. Gamma-ray spectrum from barium irradiated with

3-MeV neutrons

shown in table 25. For those elements which do not exhibit

statistically good gamma-ray peaks for peak analysis in the

spectrum, gross counts were taken over the full energy scale.

From these counts the analyst may compute upper limits to

evaluate the extent of interference in a particular situation

At the present time spectra from over seventy elements

have been recorded and their sensitivities computed. In

view of the many approximations made in this study, an error

of ±10$ has been estimated and considered adequate for the

purpose of this study.

(J. Niewodniczanski, J. E. Suddueth and S. S. Nargolwalla)

73
2

.

D-D Interference in D-T Analysis

As indicated earlier, D-D neutrons generated in a used

tritium target can present interference problems. Having

obtained the D-D spectra of the individual elements, it was

necessary only to determine experimentally the D-D neutron

build up in targets simulating the tritium loaded targets.

The only difference here is replacement of tritium by hydrogen.

Such targets are being fabricated. Some preliminary experiments with blank tantalum targets using the configuration

shown in figure 24 are in progress. A deuterium beam accelerated to 200 keV Is permitted to strike the "blank" target

and the D-D output as a function of time measured by the

neutron monitor. Using the flux normalization factor of 60

the D-D neutron build up is computed on an absolute basis.

The parameters to be studied are: 1) Acceleration voltage

effect 2) Beam intensity variations and 3) Beam characteristics such as spot size and distribution. From a knowledge of the D-D build up with time and using the computed

sensitivities, it will be possible to predict, with reasonable accuracy, expected interferences due to D-D neutrons in

any matrix during 14-MeV neutron activation analysis.

(J. Niewodniczanski , J. E. Suddueth and S. S. Nargolwalla)

3. Efficient Thermalization of 3-MeV Neutrons

Although the D-D neutron yield is about 200 times lower

than the D-T output, the thermal flux obtained by the interpositioning of a suitable moderator such as paraffin, H

p

or

D

p

between the target and the sample is relatively pure. An

experiment has been designed to make these measurements and

is shown in figure 26. The experimental arrangement has been

designed and the thermal neutron output, using light and

heavy water, will be measured. The neutron monitor location

has been changed so that the primary D-D flux can be measured

74
BIOLOGICAL

SHIELD

DEUTERIUM TARGET

ROTATING SAMPLE

ASSEMBLY

NEUTRON

DETECTOR

DRIFT TUBE

MODERATOR

(H2

or D2 0)

Figure 26. Plan view of experimental configuration used

for D-D neutron thermalization study

without interference from the moderator. Cadmium ratio measurements will assist in assessing the purity of the flux bombarding foils or liquid samples located at various distances

from the moderator. A number of moderator thicknesses will

be examined so that the optimum irradiation position can be

estimated. Information gained from these experiments will be

used for the final design of the moderator and the pure thermal

flux used for the determination of macro constituents in geological samples as well as organometallic SRMs

.

Finally, we hope to use D-D neutrons as much as possible

and reduce the analytical load on tritium targets. The

advantages in this policy are immense. Tritium targets are

generally good only for 6 or 7 mA-hours of beam time; deuterium targets, irradiated for over one hundred hours, still

provide useful fluxes. In addition, neutrons from the D-D

reaction can assist or complement the use of D-T neutrons in

the solution of complex compositional problems.

(J. Niewodniczanski, J. E. Suddueth and S. S. Nargolwalla)

75
C . Analysis of Standard Reference Materials

As was indicated in an earlier report [1], the metalloorganic SRMs are ideally suited to activation analysis with

the generator, and analytical results for four organometallic

standards subsequently certified were obtained.

The quest for higher sensitivity for oxygen in steel has

always been one of the main considerations in this project.

As will be described below, a three fold increase in the D-T

flux was realized by redesigning the target cooling cap. To

further improve matters, it is now possible to obtain sample

containers having an oxygen content about seven times less

than the "low-oxygen" containers previously used. By simple

extrapolation it appears that oxygen analyses below 10 ppm

become a definite possibility and one of the first efforts

to be made on receipt of the special containers will be to

analyze the maraging steel (SRM 109*0 for oxygen. This

material has thus far been analyzed by vacuum fusion analysis

and found to have 4 ppm of oxygen. Some of our early results

obtained for oxygen in polyethylene vials are given in this

report

.

Our experience in oxygen analysis with 14-MeV neutrons

enabled us to assist in the solution of a stoichiometric

problem associated with the preparation of high purity styrenedivinylbenzene copolymers, for the microstandard bead SRM.

Results of these analyses are given below.

(S. S. Nargolwalla)

1. Analysis of Organometallic Compounds

Four different metallo-organic SRMs were analyzed for

their metallic component. The results obtained are compared

with those from standard chemical techniques and are shown in

table 26. Attention is drawn to the excellent agreement between the two methods of analysis. The techniques employed

for the individual SRMs are described below.

76
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77
a

.

Sodium in Sodiumcyclohexanebutyrate , SRM 1069b

In this analysis thermal neutrons obtained by the thermalization of 3-MeV neutrons were used. The neutrons were

thermalized by interposing about 1.5 inches of paraffin between

the tritium target and the irradiation sample assembly. Four

samples from the same batch of this renewal SRM were analyzed

in quadruplicate. SRM 1069a (Na=11.950 was used as the

standard. The 1.37 MeV gamma-ray peak from Na decay was

counted in the multiscaler mode.

b

.

Iron in Tris ( 1-phenyl-l , 3-butanediono) iron (III)

SRM 1079b

Two samples of this material were analyzed by comparative

14-MeV neutron activation analysis using SRM 1079a (Fe=10.30$)

as the standard. Four determinations were made on each sample

(weighing approximately seven grams). A sample and a standard were simultaneously irradiated for 10 minutes and after

a 10 minute decay, counted sequentially for ten minutes each.

56 The 1.8l MeV photopeak from 2.576 hour Mn produced by the

56 56 Fe(n,p) Mn reaction was counted in the multiscaler mode.

This insured an interference-free analysis. After analysis,

each sample and its associated standard were dried according

to the procedure outlined in the Certificate of Analysis

(SRM 1079a), and the results calculated for a dry weight basis.

c

.

Copper in Bis (1-phenyl-l , 3-butanediono) Copper (II),

SRM 1080a

Quadruplicate determinations on each of the two samples

received were made using 14-MeV neutrons. Weights of all

samples were about seven grams each. SRM 1080 (Cu=l6.5%) was

used as the standard.

A sample and a standard were simultaneously irradiated

for 10 minutes and, after a twelve hour decay period, counted

sequentially for 10 minutes each with a 20 minute decay between counts. All samples were irradiated late in the evening

and counted the next morning. The 0.51 MeV annihilation radiation from the decay of Cu (T, / p=12.8 hour) was counted in

78
the multiscaler mode. The results are reported on a dry

weight basis. Drying procedure followed was that recommended

in the certificate for the standard (SRM 1080)

d. Silicon in Octaphenylcyclotetrasiloxane , SRM 1066a

Three samples of this standard were analyzed using 14-MeV

neutrons and SRM 1066 (Si=l il.l^) as the standard. Weights of

all samples were about eight grams each. Determinations were

made in quadruplicate. The three samples represented the top,

middle and bottom fractions of the bulk material. Results

indicated no detectable inhomogeneity (within the precision

of the method) and are reported on a dry weight basis. Each

sample and its standard were simultaneously irradiated for

2 minutes and after 2 minutes decay, counted sequentially

for one minute each with a one minute delay between counts.

The I.78 MeV gamma ray from the decay of the 2.31 minute

Al produced by the Si(n,p)~ Al reaction was counted in

the multiscaler mode.

(S. S. Nargolwalla, J. E. Suddueth, J. Niewodniczanski

,

E. P. Przybylowicz and Gilbert W. Smith)

2. Analysis of Oxygen in Styrene-divinylbenzene Copolymer

Four samples of this copolymer were analyzed for oxygen.

The only difference between them, insofar as a physical

examination was made, was the mesh size. Each sample was

carefully prepared in a nitrogen atmosphere and analyzed in

quadruplicate using a method previously described [19]. Each

sample was compared with a benzoic acid standard. Systematic

errors due to neutron and gamma attenuation were calculated

and found to be negligible in this case, due to the close

similarity of the sample and standard matrices. The results

are given in table 27.

7?
Table 27. Analysis of Oxygen in Styrene-divinylbenzene

Copolymer

a b Sample Identification Oxygen Concentration Uncertainty

W-121 3.46

D-103 0.134

1-11 0.37

1-22 0.44

0.07

0.006

0.01

0.01

Results are weighteu means of 4 determinations with the

weights equal to the reciprocal of estimated variances;

estimated variances based on Poisson counting statistics

and propagation of error formulas.

Errors are the standard errors of the weighted means.

(S. S. Nargolwalla, J. E. Suddueth and J. Niewodniczanski)

3 . Analysis of Oxygen in Low-Oxygen Polyethylene Vials

Low-oxygen polyethylene vials have been on the market

for about five years. These vials are made by taking appropriate precautions to avoid air oxidation of the hot plastic

during the molding process. Several laboratories, including

ours, have obtained such vials and subjected them to analysis.

The wide inconsistency of data has made it impossible to

fully evaluate the potential of polyethylene vials as carriers

for metal samples containing trace amounts of oxygen. Recently

however, we were able to obtain a few samples of a new batch

The oxygen content of these vials was determined using the

technique described earlier, and the results have been compared to those from our earlier studies in table 28, a small

batch of the new vials have been ordered and one of the first

tests will be to analyze maraging steel (SRM 1094) which

80
Table 28. Analysis of Oxygen in Low-Oxygen

Polyethylene Irradiation Capsules

Oxygen

Capsule Concentration Uncertainty

Capsule Identification Number ppm ppm13

Normal Capsules 1 373 1^

Low-oxygen Capsules 1 440 15

[Old Stock] 2 600 18

Low-oxygen Capsules 1 78 7

[New Stock] 2 87 7

Inner Liner 1 75

Polytubing

Results are weighted means of 4 determinations with the

weights equal to the reciprocal of estimated variances;

estimated variances based on Poisson counting statistics

and propagation of error formulas.

Errors are the standard errors of the weighted means.

reportedly contains only 4 ppm of oxygen. The new vials showed

marked improvement in their low-oxygen content . With a

three fold increase in flux and a better signal-to-noise ratio,

we are now in a position to analyze metal samples weighing

10 grams and containing less than 10 ppm of oxygen.

(J. E. Suddueth and S. S. Nargolwalla)

81
D. Operational Characteristics of the Neutron Generator

The operational characteristics of our neutron generator

were studied after about one year from date of installation.

Normally, these experiments are performed at the time when

the machine is operated for the first time. However, we

deliberately delayed these studies so that all components of

the accelerator had had an opportunity to age, and also to

permit us to correct any deficiencies in the system in its

early stages of beam operation. During this year about 30

mA-hour of beam was drawn using tritium targets and over 150

mA-hour using deuterium targets. Some important characteristics are described below.

a. Effect on Beam Current and Beam Diameter with

Variations in Accelerating Voltage

In neutron generator systems it is well known that the

accelerating voltage also acts as a focusing device. Briefly,

the change in diameter of the beam spot (as visually observed)

at five different beam currents was estimated as a function

of accelerating voltage. At the start of each observation a particular beam current was obtained and the focus peaked.

Then the high voltage was reduced in steps of 5 kV and the

change in spot size and beam current observed. The results

are shown in table 29.

b

.

Effect of Focus Control at 200 kV on Beam Diameter

The high voltage was fixed at 200 kV and measurements

made at five different beam currents which were obtained by

peaking the focus.- The focus control was now reduced in

small increments and the effect on the peaked beam current

and spot size observed. These results are shown in table 30.

82
Table 29. Effect of Acceleration Voltage on

Diameter of Beam Spot

Acceleration

Voltage, kV

Peaked Beam Current, mA

Approximate Diameter of Beam Spot, in

0.5 1.0 1.5 2.0 2.5

200 0.5 0.625 0.75 1.25 1.5

195 0.5 0.75 1.0 1.5

190 0.625 0.75 1.25

185 0.75 1.0 1.5

180 0.75 1.25

175 0.75 1.5

170 1.0

160 1.5

Table 30. Effect of Focus Control at 200 kV

on Diameter of Beam Spot

Focus Cont:col Beam Current Beam Diameter

(Arbitrary Units) mA in

.

54 0.5 (peaked) 0.5

^9 0.49 0.75

44 0.46 1.25

39 0.34 1.5

55 1.0 (peaked) 0.625

50 0.95 0.625

45 0.95 0.75

40 0.80 1.0

35 0.55 1.5

56 1.5 (peaked) 0.75

50 1.5 0.75

45 1.4 1.00

40 1.0 1.25

57 2.0 (peaked) 1.0

45 1.5 1.5

58 2.5 (peaked) 1.5

(J. E. Suddueth and S. S. Nargolwalla)

83
E. Health Physics

1 . Monitoring of Biological Shield

Soon after installation of the new neutron generator and

modification of the biological shield [1], a thorough surveywas carried out by the Reactor Building Health Physics staff.

The hot cell area in which the facility is located is shown

in figure 27. The total dose as measured at waist level at the

locations designated by arable letters gives the dose extrapolated for a 2.5 mA beam impinging on a new 0.5-0.8 Ci/cm tritium

target. At the console it is seen that the total dose from

gammas, slow and fast neutrons is about 27 mrem per hour.

OFFICE

CORR

CONSOLE

23

NEUTRON

GENERATOR

57

18

BI53

S

m m

BI24 CORR BI46 CORR

BI52

BI47

Figure 27. Location of radiation monitoring sites around

the neutron generator facility

84
It is therefore possible to operate the accelerator for about

3 hours a week at full beam on a new target . Since the target

depreciates by a factor of two every 2 mA-hours, permission

has been granted by the Health Physics staff for operation

times up to 10 hours per week. In the context of 1^-MeV

neutron activation analysis this is a very large amount of

beam time using tritium targets. With deuterium targets,

however, it is possible to operate over forty hours per week.

2 . Tritium Contamination and Control

Tritium contamination problems associated with all D-T

type accelerators were examined in great detail after 1 1/2

years operation of the accelerator. During this period of

time about 30 mA-hours of beam was used on a total of four

tritium targets. The accelerator was subjected to a thorough

check for tritium levels inside various components integral

to the accelerator and inside subsidiary systems.

A summary of this investigation is shown in table 31. The

exact locations at which sampling was done are designated and

shown in figure 28. Despite the extreme care we have taken in

ACCELERATOR SECTION

SOURCE /®T PUM

DRIFT

C00LING LINES

(?>h\

® W<©

/TUBE

VACUUM

HOSE'

TO

^D

ROUGHING

FREON |-| VACUUM

STORAGE PUMP

TANK % (S^PUMP

OIL

J-TARGET

--(D FREON

COOLING

TANK

-v®

5>

FREON

CIRCULATING

PUMP

Figure 28. Location of tritium monitoring

neutron generator

sites on the

85
Table 31. Tritium Contamination of the Neutron Generator

After 30 mA-hrs . of Beam Operation on four

0.8 Ci/cm2 Tritium Targets (^ 1 1/2 Years Operation)

Component

Monitored

Roughing Pump System

(a) Pump il

(b) Roughing Pump Hose

(i) External

(ii) Internal

(pump end)

(iii) Internal

(Generator End)

Monitoring

Location

(Fig. 6)

Circulating Cooling Pump

(a) Inside Seal

(b) Inside Pipe Threads

3. Freon Cooling Tank

4. Plastic Container-Freon

Storage

5. Ion Source

(a) Inside

(b) Electrodes

6. Drift Tube

3

6

5

6

7

8

9

10

(a) It Inside Ion Pump

Connection 11

(b) Inside at Ion Pump

Flange 12

7. Ion Pump

(a) 0-ring (Front of Pump) 13

(b) 0-ring (Rear of Pump) iH

Tritium

Contamination

O.lyCi/ml

_5 1x10 uCi/swipe

_?

1x10 uCi/swipe

6x10 J uCi/swipe

_i|

7x15 uCi/swipe

5x10 yCi/swipe

0.2 uCi/swipe

-2

3x10 uCi/swipe

2x10 3 yCi/swipe

7x10 yCi/swipe

0.1 yCi/swipe

-2

7x10 yCi/swipe

4x10 ~yCi/swipe

1x10 yCi/swipe

86
the installation of tritium targets some degree of contamination has always existed in our target assembly hood. The

Health Physics staff are now monitoring this hood daily. At

the accelerator site, however, tritium contamination of the

outer surfaces of the accelerator is negligible.

3. Bremsstrahlung Dose Measurements

When positive ions are accelerated onto a target, the

emitted electrons are accelerated in a reverse direction.

In a generator these electrons are suppressed, however a great

majority are accelerated back to the terminal end of the

accelerator and generate bremsstrahlung. The maximum energy

of the bremsstrahlung is 200 keV and present a hazard even

when protons are accelerated on to a blank target. The

extent of this hazard can be assessed from the results given

in table 32 with the aid of figure 29 which shows the locations

of dose measurements.

HIGH

VOLTAGE ION

DOME VACUUM

PUMP

@ TARGET

DRIFT TUBE

Figure 29. Bremsstrahlung dose measurement sites around

the neutron generator

87
Table 32. Bremsstrahlung Dose Measurements as a Function of

Beam Current (Hydrogen on Blank Cu Target)

at 200 kV Acceleration Voltage

Dose mrem/hr

at Contact

Location

Monitore 3 Beam Current

(Fig 7)

Voltage

mA

1. High

Terminal Dome

(a) Back (1) 0.5

(b) Side (2) 0.5

(c) Bottom (3) 0.5

(a) Back 1.0

(b) Side 1.0

(c) Bottom 1.0

(a) Back 1.5

(b) Side 1.5

(c) Bottom 1.5

(a) Back 1.75

(b) Side 1.75

(c) Bottom 1.75

15

100

20

40

310

50

80

1750

100

175

2500

200

Target End

Between Suppressor Beam Current Dose

and Cooling Cap (4 ) mA mrem per hour

0.5 0.6

1.0 1.4

1.5 2.8

1.7 3.0

2.0 5.8

2.35 9.5

(J. E. Suddueth and S. S. Nargolwalla)

88
p. Status of the Neutron Generator Facility

For the first year after its installation, the neutron

generator performed flawlessly. After this, certain key

components began to fail causing significant amounts of downtime for repair. Some of the more serious breakdowns and the

measures taken to reduce downtimes are described. During this

period of time inovations were also made to insure safety of

operating personnel and the generator; and also to increase

the available D-T flux by a factor of about three. At the

present time we are in the process of becoming self-sufficient

in terms of fabricating key components in our shops and obtaining knowledge necessary for reconditioning of other components in our laboratory. The major part of the delay has

been in the transportation of the defective unit to and from

the manufacturer. In the near future we hope that this selfsufficiency approach will prevent breakdown and also reduce

downtimes to negligible proportions.

1

.

Vacuum Ion Pump Failure

A direct short in the pumping elements necessitated the

removal of the pump for installation of new elements. Due to

the large tritium load in the pump, difficulties were experienced in its transportation from the laboratory. The ion

pump was returned after six weeks and found to be working at

very low efficiency. Since then, however, the manufacturer

has shipped a new pump. Experiments carried out showed that

the pump, when pumping on itself, could not reduce the vacuum

_7 below 10 mm Hg . This was unacceptable to us, as the pump

-9 when installed was capable of pumping down to 10 mm Hg

.

It is suspected that there may be a leak in the pump housing

itself.

2

.

Ion Source Burnout

After a very good performance for over a year, the ion

source finally burned out . Insofar as this type of source

is concerned, this performance is difficult to exceed. On

89
dissassembly, it was discovered that the anode was extensively

burned. The dissassembled source is shown in figure 30. From

left to right are the mechanical gas leak attached to the

source housing, burned anode, retaining sleeve, front cathode

Figure 30. Penning ion source: disassembled

and assembly holder. In figures 31 and 32. the burned anode

shown indicates the extent of the damage. Despite this damage,

the source was still operating at about 40% of its rated

capacity; a fact which speaks highly for the Penning type

source. In figure 33 the location of the back cathode and

the anode electrode is shown. The assembled source with

magnets is shown in figure 3^.

/;ji/ ^.j.jii

1 ili^j|i|^

;

{ ,V

v

Figure 3.1. Penning ion source: inside view of burned

aluminum anode

90
-^mmtr* '

'/' 'i'l'j .' 'TlM't'l 1 *! *5

Figure 32. Penning ion source

aluminum anode

side view of burned

Figure 33. Penning ion source: inside view of back

cathode and protruding anode head

Figure 3^. Penning ion source: assembled, showing

mechanical gas lead, and U magnets

91
These sources are extremely rugged in construction and

require little attention. The art apparently lies in the

replacement of anodes and cathodes. The back cathode, however,

requires a special tool and great care in its installation.

Although a spare source was purchased, we are now in the

process of reconditioning ion sources ourselves.

3 . High Voltage Breakdown

Recently, arcing in the main high voltage tank was experienced. A careful examination which included a thorough

cleaning of the oil and all components did not reveal the

cause. However, a closer examination of the heavy polyethylene covered high voltage cable showed a pin hole in which

carbon deposits could be seen. The cable was cut off at

this point . The defective piece was carefully trimmed and

showed multiple discharges via carbon tracks to the 1/4 in.

diameter copper conductor. A photograph illustrating this

discharge pattern is shown in figure 35.

(f ft h'ii,

-: ; ,a

,7.7 //.yV7wy'/'/'/7';'^'/'l',"l'j'i'r4!

j

, l'I'l'| M'iMj'j'i'lMT r'; ;> ;

Figure 35. 200kV conductor showing electrical break-down

of insulator

92
4 . Cooling Problems

The use of Freon as a coolant required the use of a specially designed cooling cap shown in figure 36. Although

cooling is extremely efficient, the design of this cap was

such that the minimum distance a sample could be placed from

the target was almost 1.5 cm. We have now changed over to

water cooling and employed a thin stainless steel cooling

cap and have realized a factor of three in flux. The minimum

distance from target-to-sample is less than 1 cm. The exact

distance is difficult to determine since the target itself

bulges inward under vacuum and from the water pressure in the

cooling cap. It was also discovered that with the Preon

cooling cap, tantalum blank targets could not be used since

they did not bulge inward being much stiffer than the copper

backing plate of the tritium target. Due to this fact the

cooling cap would make contact with the blank tantalum plate

with loss of cooling surface. In figures 37 and 38, the beam

side and cooling side of a typical tantalum plate in which a hole has been burned through by the deuterium beam are shown.

Figure 36. Target cooling jacket used with Preon

93
i'l'I'l'l'l'IT'^ iri*\4

Figure 37. Beam-side of blank tantalum target showing

burned hole

I

"

Figure 38. Coolant-side of blank tantalum target showing

burned hole

5 . Safety Devices

About one year ago a serious accident was reported in another accelerator facility. An investigation into this event

indicated short comings in the interlock systems employed.

Since then we have re-examined our system and added to it

.

Previously it was possible to turn on the beam and not know

whether anybody was inside the irradiation area. Our double

interlock system makes it mandatory to check inside once the

interlock door has been opened for any reason whatsoever.

94
To protect the facility efficiently, we have installed

thermocouples in the coolant input and output lines with

external readouts. The heating effects can be monitored continuously during beam operation. At a flow rate of about

0.5 gallons per minute, the temperature rise from the ambient

is about 20°F at 2.5 mA beam current. The refrigeration

system controls the coolant input temperature to about 60°F

over long periods of beam time. If the refrigerator is shut

down, the tank water supply temperature increases at the

rate of about 1°F per minute. On the basis of this it is

still possible to run the accelerator up to thirty minutes

should the refrigerator fail. In view of our efforts to

upgrade the facility and improve Its performance and reliability we expect insignificant downtimes in our future operations . (J. E. Suddueth, S. S. Nargolwalla and J. Nie wodniczanski)

95
5. ACTIVATION ANALYSIS WITH THE NBS LINAC

A. Introduction

The primary emphasis in the LINAC project for this report

year has been the development of procedures for the trace

determination of carbon in a variety of metal samples, more

work on oxygen and carbon in sodium and the development of

methods for the determination of thallium at trace levels in

simulated lunar samples. In addition, some preliminary

experiments have been performed for the determination of

bismuth by photof ission

.

B. Facilities

The second rabbit terminal installed in the 45 degree

irradiation facility behind the original terminal continues

to function in a satisfactory manner. This rabbit terminal

has been used for the production of isotopes for Mossbauer

spectroscopy and radiochemical separations as well as for

preliminary experiments in activation analysis. Figure 39

shows a portion of the control console. The control lights

indicate a sample in both irradiation positions . A view of

the LINAC activation analysis laboratory in Service Area 2

at the Radiation Physics Building is shown in figure 40.

•

^f

&-

JW'LIS

Figure 39. LINAC Control Console

96
Figure 40. LINAC Activation Analysis Laboratory in the

Radiation Physics Building

The zinc sulfide phosphor previously described [1] continues to function well and is of assistance to the accelerator

operators in beam positioning. Only very recently has the

phosphor deteriorated and been replaced.

1 . Carbon in High Purity Metals

A substantial effort to develop methods for determination

of carbon in a variety of high purity metals has been initiated

Preliminary experimentation with perchloric acid oxidation was

not successful as yields were low and not reproducible.

A resistance heated furnace was obtained and separations

were made using "classical" procedures with modification when

appropriate. The samples were irradiated in metal rabbits

and etched or ultrasonically cleaned after irradiation.

97
Irradiations were monitored with a small piece of graphite in

the lid of the metal rabbit. Standards were spectrographic

grade graphite. Counting was done with the coincidence unit

described in the sodium section. The resistance furnace and

collecting train is shown in figure 4l.

Figure 41. Resistance furnace and collecting train for

carbon determination

As a check on the method, samples of SRM 131a (iron) which

is certified at 44 ppm were analyzed. Values of 45, 48, 45,

43, 40 and 46 ppm were obtained.

Three Standard Reference Materials; 680 (platinum), 682

(zinc) and 685 (gold) were also analyzed for carbon content.

Gold . The gold samples, weighing about six grams each,

were etched after irradiation in aqua regia to remove about

3% of the sample. Approximately four grams of lead flux was

98
added and the extraction was carried out at 1200°C with an

oxygen flow rate of 500 ml/min. Extraction times were approximately fifteen minutes. Values obtained were 0.44, 0.39 and

0.49 ppm. The decay curve associated with one of these samples,

is shown in figure 42.

40 60 TIME (MINUTES)

100

Figure 42. Decay curve for carbon-11

99
Zinc . The zinc samples, weighing about ten grams each,

were etched after irradiation in dilute hydrochloric acid to

remove about 20% of the samples. No flux was used. The

combustion was carried out at 950-1000°C with an oxygen flow

rate of 500 ml/min. The oxygen was moistened by passing it

over 1M HpSOj. to control the combustion. The extractions were

carried out in a transparent quartz tube so the course of the

reaction could be observed. After the burning had subsided

(typically 10-15 minutes), the system was allowed to flush a few minutes before counting. Inspection of samples after

burning indicated complete combustion. Values obtained were

0.3, 0.2 and 0.2 ppm.

Platinum . The samples, weighing between 0.5 and 2 grams

each, were etched after irradiation in aqua regia to remove

varying amounts of sample, or were cleaned ultrasonically

before extraction. Approximately four grams of lead flux

were added and the extraction carried out at 1200°C with an

oxygen flow of 500 ml/min. Extraction time was approximately

15 minutes. Table 33 shows values of apparent concentration

of carbon as a function of amount of sample removed by etching from the surface before separation.

Table 33. Effect of Etching on Apparent Carbon Concentration

Amount of

Sample Removed Carbon Found (ppm )

17, 24

*1% 6 .

5

9% 0.5

A value of 27 ppm also obtained with 9% etching was rejected as obviously not in line with the rest of the results.

The 0.5 ppm value has a large uncertainty associated with it

and probably cannot be considered different from the 1 ppm

value

.

100
2. Thallium in SRM Glass

The Office of Standard Reference Materials has obtained a

series of glass samples which have been doped with 6l trace

elements at levels of 500 ppm, 50 ppm, 1 ppm and 20 ppb . We

are investigating the determination of thallium in these

glasses by photon activation analysis.

The nuclear reaction involved in these determinations is

203 202 Tl(y,n) Tl. The product nuclide decays with a half-life

of 12 days by electron capture with the emission of the 70

keV mercury X-ray in coincidence with a 439 keV gamma ray.

The cross section for (y,n) reactions increases and the threshold decreases with increasing atomic number; thus the intrinsic

sensitivity for thallium, with the 203 isotope having a 29.50%

abundance, is good.

Since a substantial amount of activity from the other 60

elements, as well as the major constituents of the glass, are

present after irradiation, a separation is required. From

work on tetracycline extractions reported elsewhere in this

report (Part D) it was determined that thallium (III) could

be extracted in a wide pH range with benzyl alcohol.

A sample of the 50 ppm glass was obtained and about one

gram irradiated in the 45 degree facility of the NBS LINAC for

about 45 minutes.

The sample was dissolved after irradiation in a mixture

of 10 ml of perchloric and a few ml of hydrofluoric acids

with heating. To insure the presence of thallium (III), 10

ml of HC1 was added to the perchloric acid with 10-12 drops

of hydrogen peroxide. Excess peroxide was expelled by heating

and the solution was siluted to 40 ml with distilled water.

The purpose of the dilution was to prevent concentrated perchloric acid from destroying the benzyl alcohol. This solution

was extracted twice with ten ml portions of benzyl alcohol for

several minutes. After each extraction, the separatory funnel

was centrifuged for a few minutes to facilitate phase separation. The organic layer was washed three times with a mixture

101
of 10 ml HCIO^, 10 ml HC1 and 20 ml H

2

0. Centrifugation was

also necessary after each wash. The organic phase was then

drawn into a two oz. polyethylene bottle for counting. Tracer

experiments showed that the separation was quantitative.

The organic phase was counted with a Ge(Li) detector and

2048 channels of analyzer memory. Gamma ray lines that could

be ascribed definitely to gold activities and tentatively

to antimony activities were observed in addition to the thallium activity.

3. Photofission of Tl, Bi and Pb

The nuclear track technique has been a method predominently used to study radiation induced trails in plastic solid

state detectors by thermal neutron irradiations [20-24],

Currently, investigations are underway to study the determination of several heavy elements by counting tracks produced

by photofission. Fission tracks have been produced in Lexan*

polycarbonate from photofission of thallium, lead, bismuth,

thorium, and uranium produced by 35 MeV photon irradiation.

However, five minute irradiations for one microgram samples

indicated that only the latter three elements should be considered for quantitative measurements.

Since trace determination of bismuth is difficult by

neutron activation analysis and thorium and uranium are relatively easy, Bi appeared to be the most promising element

for initial attention even though sensitivity for bismuth,

_7 3.13x10 grams per track, was found to be less than those

i -9 for thorium and uranium, 4.17x10 grams per track and 1.72x

-9

10 grams per track respectively. Since the uranium and

thorium sensitivities were greater than that of bismuth, a pre-irradiation separation had to be performed on the material

to be analyzed.

*See disclaimer in Preface.

102
a. Experimental

The bismuth containing material was dissolved in dilute

HNO~ and the solution mixed with 10 ml of a mixture containing

50 grams of Na

?

EDTA and 50 grams of NaCN dissolved in 1 liter

of 1.5N NH.OH. One milliliter of a 2% solution of sodium

diethydithiocarbamate was added and the bismuth extracted into

10 ml of CCl^ [25]. The CCl^ layer was drawn off, 1 ml of

dilute HN0~ was added and the mixture heated until all of

CCIk had evaporated. After diluting to 10 ml, 100 ul of the

dilute HNO^ solution was placed on Lexan slides^ 10 mm wide x 25 mm long x 0.25 mm thick. The solution was then evaporated,

covered with collodion, and irradiated for a minimum of five

minutes with bremsstrahlung produced by 35 MeV electrons.

After irradiation the slides were etched in 6.5N NaOH at 50°±

2°C for 45 minutes to develop the tracks. The tracks were

then counted and values were compared with values obtained from

standards irradiated simultaneously.

b

.

Conclusions

Analysis on both fractions of solvent extraction layers

by the nuclear track technique showed that all the uranium

remaining in the aqueous phase. Tracer experiments with 207 Bi

and the track technique indicated that 95$ of the bismuth was

being extracted into the CCl^. These results indicate that

trace quantities of bismuth and uranium can be analyzed in

the same matrix.

4. Yttrium in Rare Earths

The photonuclear activation analysis technique for the

determination of yttrium without chemical separation described

in last year's report [1] has been applied to the determination

of yttrium in several rare earth oxides. The determination of

yttrium is difficult using thermal neutron activation analysis

because the product nucleus decays exclusively by beta emission

and must be separated from chemically similar elements.

103
O q

o o We use the nuclear reaction y Y(y,n) Y for this determio o nation. Y has a half-life of 108 days and decays by emission of 898 keV and 1827 keV gamma rays in coincidence.

The threshold for the reaction is 12.0 MeV, the energy of

maximum cross section is 16.3 MeV, the value of this maximum

cross section is 191 millibarns and the width at half-height

of the cross section curve is 3-8 MeV.

Some results obtained on rare earth oxides are given in

table 3^.

Table 34. Results of Yttrium Determination for

Some Rare Earth Oxides

Sample Y 2°3* (ppm)

Nd

2

U

3

16,17,17

Pr

2 3

421,408,430

Sm

2

17,19

Gd

2 2

26,25

Tb

2 171,180

Dy

2 3

175,165,176

Two potential sources of systematic error were investigated to assure that they did not interfere with the determination. At least one of each of the rare earth samples in

the table was counted two or three times over a period of

several months and compared with its standard. The apparent

concentration of yttrium did not change and one may infer

from this that there are no interferences in counting.

The possibility of differences in self-shielding during

counting was examined by spiking inactive rare earth oxide

00

and yttrium oxide samples with known amounts of Y. These

samples were made into the configuration of the counting

104
samples and counted under identical conditions. The ratio of

on counting rates was within one percent of the ratio of Y

added.

For the general determination of yttrium in a variety of

materials by this nuclear reaction, a cursory examination of

the problem of interfering nuclear reactions was made.

The potential interfering reactions are

93 Nb(y,an) 88

Y

Zr(y,np) Y

Zr(y,2n) Zr -> Y

Samples of niobium, zirconium and yttrium were irradiated

at 35 MeV and the relative amount of flux received by each

sample monitored with copper foils. A few days after irradiation the niobium sample was examined with a Ge(Li) detector.

No Y was observed and an upper yield limit of about 10

for the production from niobium relative to that from yttrium

on an equal weight basis was obtained. The zirconium sample

was examined a few hours after irradiation and again after

several weeks. A few hours after irradiation, a small amount

o o of Y was observed and an equivalent upper yield limit of

— il P P about 10 was obtained. Several weeks later, Y had grown

on in from Zr to such an extent that the count rate relative

to the monitor and weights of zirconium and yttrium was of

the order of one percent of the yttrium. Thus if zirconium

is known or suspected to be present in a given sample, it is

necessary to separate the Y from Zr if the sample is not to

be counted promptly.

5 . Carbon and Oxygen in Sodium

A substantial effort has continued on the determination

of carbon and oxygen in sodium. We have cooperated and are

discussing cooperation with other laboratories on the comparison of results from samples drawn from the same lot.

105
In photon activation analysis the nuclear reactions and

half-lives of the products of interest are

12„ f xllp C(y,n) C

l6 n , a5n 0(y,n) y

20.5 minutes

2.1 minutes

Both radionuclides decay by positron emission. Their halflives are of sufficient length to permit a simple separation

There are two major advantages in accomplishing these

analyses by photonuclear activation analysis. These advantages are frequently available in most activation analysis.

One is the ability to etch the sample after irradiation, but

before separation and counting, thus eliminating reagent

blanks and the possibility of contamination of the sample

thereby obtaining an analytical result truly representative

of the bulk sample. Our samples are etched in a solution

of ethyl alcohol containing about ten percent water. The

etching process is shown in figure 43. It is apparent

Figure 43. Etching of sodium in alcohol-water mixture

106
from the figure that the evolution of hydrogen bubbles assures

the removal of surface impurities from the sample. -Generally,

three or four such etches are made for each sample. In the

case of oxygen, if there is a substantial amount of sodium

hydroxide and sodium carbonate on the surface, it is useful

to etch prior to irradiation so that the post irradiation

etch does not require excessive time.

The second advantage of activation analysis is that the

final "analytical signal," the counting of the separated

activity, decays at a rate characteristic of the nuclide

produced. Since many of the isotopes produced by photon

activation decay by positron emission, separations for the

element of interest must be clean as well as be completed

in a time commensurate with their half-lives. They also must

be quantitative or amenable to yield determination.

The separation of carbon has been described [26]. The

technique utilizes the "low temperature combustion" method.

The apparatus is shown in figure 44. Oxygen enters the

assembly via the three-way stopcock at the top center of the

photograph. The irradiated and etched sodium sample with

Figure 44. Combustion of sodium for carbon determination

107
about 100 mg of inactive sodium carbonate in a quartz crucible

is heated with a gas burner. When the sodium melts, the

heating is stopped and the stopcock directly below the

Kjeldahl bulb is closed. One to two seconds later combustion

occurs. The sodium is burned to sodium oxide and the carbon

is converted to sodium carbonate. As seen in the figure,

sodium oxide diffuses above the combustion. The closed stopcock prevents the oxide from coming in contact with the

carbon dioxide liberated later. When the crucible has cooled,

acid is added via the three way stopcock to liberate the

radioactive carbon as COp. Boiling under vacuum refluxes the

acid solution throughout the combustion vessel and neutralizes

sodium oxide which has diffused to the walls. The exit gas

is scrubbed in a solution of sulfuric and boric acids to

nq -I p

remove radiof luorine produced by the reaction Na(y a an) F.

The C0

2

is trapped in a sodium hydroxide solution. The activity in this solution is counted by measuring the 511 keV

radiation from position annihilation. The two annihilation

quanta are emitted at 180° angles and are detected with two

4 in. x 4 in. Nal(Tl) detectors and a coincidence apparatus.

The coincidence unit, scaler and pulse height analyzer are

shown in figure 45. Background with this detection apparatus is typically 0.2 cpm.

Figure 45- Coincidence unit for positron annihilation

counting

108
The apparatus for the separation of oxygen in sodium is

shown in figure 46. With joint J open, irradiated and

etched sodium is dropped into distillation flask A. The flask

contains 25 ml of water in which six grams of sodium hydroxide

has been dissolved to temper the reaction rate of the sodium

with the water. Nitrogen introduced via stopcock SI flushes

the hydrogen from the system.

Figure 46. Apparatus for separatjnn of oxygen from sodium

The oxygen, present as Na ? 0* after activation, is converted to sodium hydroxide upon dissolution and the radioactive oxygen exchanges with oxygen in the water. After

dissolution, joint J is connected, stopcock S2 is closed and

stopcock 2 is opened to vacuum. The distillation flask is

heated with a blast burner and about one half of the water

is distilled into the liquid nitrogen cooled trap. After

distillation, the quantity of water distilled is measured,

diluted to a constant volume and the annihilation radiation

counted. Some results are shown in table 35.

Table 35.

Reagent Grade

Sodium - Lot I

Reagent Grade

Sodium - Lot II

Reagent Grade

Sodium - Lot III

Sodium Correction

Loop Experiment

Oxygen in Sodium Results

90,77,82 ppm

51,42,47 ppm

100,105,110,100 ppm

29,22,20 ppm

109
6. RADIOCHEMICAL SEPARATIONS

A. The Use of Hydrated Antimony Pentoxide (HAP) for

Separation of Elements in Irradiated Materials

1 . Introduction

In the neutron irradiation of many materials, such as

those of biologic origin, geological specimens, glass, etc.,

for which activation analysis is a useful tool, the large

amounts of sodium present result in very high levels of gamma

activity. This activity can cause a radiological health

hazard, as well as serious problems in the analysis of the

gamma spectra for other nuclides present.

Girardi and co-workers at ISPRA [12] reported the use of

hydrated antimony pentoxide (HAP) for the selective, quantitative removal of radiosodium from mixtures of radionuclides

.

A study of the use of HAP as applied to the NBS activation

analysis program was started by attempting to repeat some of

the work reported by Girardi. First, the retention of sodium22 in 12M HC1 on HAP was measured. Quantitative removal was

observed. However, difficulty in the handling of the HAP

was encountered. The HAP proved to be very fragile, i.e.,

it continually broke into extremely fine particles, even

after sieving and slurrying with HC1 or water to remove the

"fines ."

The HAP was used 'in 11mm diameter columns . If the filtering and supporting media at the bottom of the HAP column were

not retentive enough some of the "fines" came through with

the eluant, carrying appreciable quantities of the sodium

along. If the filtering and supporting media were too retentive, the "fines" plugged it up and little or no flow through

the column resulted. After trials with quartz-wool plugs' as

recommended by Girardi gave very poor results, course fritted

filter discs, selected by trial and error were found to be

satisfactory. Figure 47 shows a column found to be useful for

quantities of HAP ' up to 9 grams and solutions up to 50 ml.

110
Figure 47. Sodium removal on a HAP column

2 . Experimental

??

a. Retention of ~~Na on HAP from 12M HC1 . One half

ml of ~~Na solution and 15 ml 12M HC1, with from to 70 mg

NaCl as carrier were added to a 2 oz. polyethylene screw-top

bottle, and counted on a 3 in. x 3 in. Nal(Tl) crystal connected to a 1024 channel analyzer. This was the standard

counting procedure for all samples

.

The solution was then passed through a 3 gm column of

HAP prepared as above, and the eluant collected in a clear

2 oz. polyethylene bottle and counted. The flow of the solution through the HAP column was assisted as necessary by

light air pressure applied to the top of the column. Flow

rate was held to 2-3 ml/min. In all cases, retention of

22 the Na on the HAP column was quantitative. The columns

were then washed with two successive 15 ml portions of 12M

HC1, and the eluates caught and counted separately. In all

22

cases, only small fractions of a percent of the Na showed

in the eluates.

Ill
b. Separation of 2Na and 110m

Ag in 12M HC1 Solution . 22 Two hundred and fifty yl of the Na solution, 0.5 ml of a

solution of 110mAg and 15 ml 12M HC1 were added to a 2 oz

polyethylene bottle and counted. The solution was then

passed through a 3 g column of HAP, the eluate collected and

22 counted as above. The Na was quantitatively retained on

the HAP, and 92% of the Ag was present in the eluate.

The HAP column was rinsed with 15 ml 12M HC1, which was col22 lected and counted as above. A trace of Na and the remainder

of the Ag were present. The HAP column was rinsed with a

22 second 15 ml portion of 12M HC1. Again, a trace of Na was

present

.

22 110m

c

.

Separation of Na and Ag from Solutions of HC10 f| Since many of the samples analyzed in the Section's program

are processed in HClOj., a study of the behavior of HClOj.

solutions of various radionuclides on HAP columns was initiated.

The ~~Na solution containing Ag and ^50 mg NaCl in 15 ml

of 12M, 6M or 3M HC1CK respectively were counted.

d. 12M HCIO l . The solution was passed through a 3 gm HAP

column and the eluate was collected and counted as above.

Approximately 6Q% of the "Na and no Ag were present in

the first eluate. The HAP column was rinsed with 15 ml 12M

HCIO^. Eluate #2 showed ^32$ of the original ~"Na present,

and no Ag. The HAP column was rinsed with a second 15

ml 12M HCIOk. Eluate #3 showed <1% of the original Na and

-] -I A no Ag. These results were extremely interesting, since

Girardi had reported that he was unable to remove sodium from

HAP with any of the acids and solvents that he had tried.

e. 6M HC10 ||. The solution was passed through a 3 gm HAP

column and the eluate collected and counted as above. No

~~Na or Ag were detected. The HAP column was rinsed with

two separate 15 ml portions of 6M HC10,., and the eluates

22 collected and counted separately. No Na was detected in

either eluate, and only trace amounts of Ag were observed.

112
The HAP column was then rinsed with 15 ml of 12M HC1. The

?? _ 110m eluate contained no ~~Na and ^15$ of the Ag. The HAP

column was then rinsed with a second 15 ml of 12M HCl, and

the eluate ^2.5$ of the 110mAg and ^3% of the 22

Na were

present

.

f. 3M HCIO k. The solution was passed through a 3g HAP

column. No detectable amounts of "Na or Ag were in the

eluate. The HAP column was rinsed with 15 ml of 3M HCIO^.

22 110m The eluate again showed no Na or Ag activities. The HAP

column was rinsed with 15 ml 12M HC1. The eluate contained no

~~Na but ^70% of the m

Ag was present. It was attempted to

rinse the column with a second 15 ml portion of 12M HCl, but

the column clogged and only about 10 ml could be passed

22 through. The eluate contained no Na, but significant

amounts of Ag, which could not be measured quantitatively.

In further studies, after refinement of the HAP column

11 Hm

preparation procedure, it was shown that the "Na and Ag

could be recovered sequentially and quantitatively from HAP

columns on which they had been deposited from 3M HCIO^

solutions, using successive eluations with 12M HClOj. and 12M

HCl. However, varying amounts of the Ag, from zero to

15% came through with the original 3M HClOj. fraction.

198 110m 22

g. Separation of Au, Ag and Na with HAP . Using

a solution of 19 Au,

110m

Ag and 22Na in 3M HCIO^, and the

eluation procedures described in (a) above, it was found that

the gold came through quantitatively in the first two 3M

HClOn eluates, 9&% in the first and the rest in the second

110m 22 eluate. The Ag and Na were then separated by successive

eluations with 12M HCl and 12M HCIO^ as described in (c)

above

.

h. Separation of Radioactive Sodium from Glass . One

hundred and forty mg of MG GCA-41 glass (not irradiated) and

22

. 5 ml Na solution were dissolved in HF, HNCK and HClCv,

taken to dense fumes of HCIO^., and, very carefully, to dryness. The residue was then dissolved in 12M HCl and counted.

113
The solution was passed through a 3g HAP column, and the

eluate collected and counted as above. Approximately 0.3$ of

22 the Na showed up in the eluate.

A number of such HAP columns were then prepared and given

24

to Dr. Barbara Thompson, who used them to separate Na from

solutions of irradiated glass samples

.

Two aliquots of a solution of irradiated glass 40-B, obtained from W. D. Kinard, were diluted to 15 ml with 3M HC10|.,

and passed through separate 3g HAP columns. Each column was

rinsed with two 15 ml portions of 3M HOICK, which were collected and counted separately. The sodium was retained

quantitatively on the HAP columns. Gold-198 came through

quantitativley in the first eluate. Other radionuclides

present, not completely identified, came through the first

and second eluates in approximately equal proportions, and in

lesser, but significant amounts in the third eluate. Prom

this preliminary work it would appear that 3M HCIO^. is not

the ideal medium for initial passage of a complex mixture of

radionuclides through HAP. A medium which would allow either

quantitative passage of all or most of the radionuclides

through the HAP, or which would permit quantitative retention

of all or most of the radionuclides on the HAP would be most

desirable. The former exists, e.g., 12M HC1 as reported by

Girardi . The latter will be a goal of further investigation

here, along with means of selectively eluting the absorbed

radionuclides

.

3- Results and Discussion

Studies have shown that hydrated antimony pentoxide (HAP)

as reported by Girardi and Sabbioni Ql2 ] at ISPRA, is very

effective in removing sodium from solutions. Further, these

preliminary studies have shown that by use of suitable solvents, namely different acids of various concentrations, with

HAP, selective absorption and elution of a mixture of radionuclides may be effected. Further investigative efforts

114
along these lines will be carried out, possibly in conjunction with studies of other inorganic ion exchangers, such

as SnOp, which have been rennr-ted in the literature.

(W. F. Marlow)

B. The Application of the Antibiotic, Tetracycline , to Chemical Separations

1. Introduction

The group of antibiotics called the tetracyclines have

found wide application in the field of medicine since the

early 1950' s. The group includes such antibiotics as tetracycline (achromycin), oxytetracycline (terramycin) , and

chlorotetracycline (aureomycin) . Tetracycline is the generic

name for 4-dimethyl amino-1, 4 , 4a, 5 , 5a, 6 ,11 ,12a-octahydro-3

,

6 ,10, 12 , 12a-pentahydroxy-6-methyl-l, 11 dioxo-2 naphthacene

carboxamide. It is a yellow crystalline compound having the

formula C 22 H 2ij N

2 0g and a molecular weight of 444.43. The

structure of the tetracycline molecule is shown in figure 48.

R, CH 3 OH

I?

R

B

CH 3 ( +) CH,

/

Figure 48. Structure of the tetracyclines

Oxytetracycline R =H, Rp=0H

Cyc lot etracyc line R-, = C1,Rq= H

Tetracycline

115

R-. — H, R^-)—

H
The groups A, B, and C illustrate the three distinct

acid groups of the molecule, the tricarbonylmethane system A,

the ammonium cation B, and the phenolic diketone system C.

The three thermodynamic pka values have been assigned by

Leeson [27] et at. as pK, (system A)=3.30, pK„ (system C)=7.68,

and pK (system B)=9.69-

Tetracycline has been recognized for some years as an

excellent chelating agent for various inorganic ions. Its

action as an antibiotic has been attributed to its strong

chelating properties [28].

Albert [29,30] has calculated stability constants for the

interaction of many metal ions with tetracycline .HC1, oxytetracycline .HC1, and chlorotetracycline .HC1, and has found

many to form stable complexes. Maxwell et al. [31] have determined the stability constants of magnesium, calcium, and

strontium with tetracycline. Other studies have postulated

that tetracycline facilitates the elimination of strontium

from bones in mice [32,33], although Maxwell, et al. dispute

that this enhancement is due to a strontium-tetracycline

complex since calcium forms a more stable complex with tetracycline than does strontium.

Ibsen and Urist [3*0 have postulated the positions of

chelation of the molecule with calcium to be between the

oxygens at carbon positions 10 and 11, 11 and 12, the oxygen

at carbon 1 and the NH ? at carbon 2., and the oxygens at

carbons 2 and 3. Ishidate and Sakaguchi [35] have studied the

chlorotetracycline complexes of various metals in order to

predict the point of chelation in the molecule. Baker and

Brown [36] have concluded that chelation with cobalt and

nickel takes place through two oxygens of the 1,2,3 tricarbonmethane system, the amide oxygen at C

?

and the hydroxyl

at C-. or C^.

Although much work has been done with tetracycline in

structural determinations of stability constants, no one had

116
previously applied tetracycline to chemical separations,

although one would postulate from the structural studies previously mentioned that it may be useful in that capacity.

The first application of tetracycline for an analytical

separation was a solvent extraction produced for the determination of strontium-90 in environmental grass samples [37].

A study was then conducted to determine the feasibility

of a comprehensive examination of the extraction of the

metallic elements using tetracycline. The metals chosen for

this preliminary study were titanium, scandium, iron, nickel,

yttrium, niobium, thallium, and lead.

2 . Experimental

The tetracycline used in this study was supplied by

Charles Pfizer and Co.* as the hydrochloride. Benzyl alcohol

was chosen as the organic solvent because it readily dissolves

the tetracycline and is immiscible with water. In preparing

the solution the tetracycline-benzyl alcohol mixture was

heated slightly to facilitate solution and was cooled to room

temperature before use in the extraction studies

.

The procedure used in this study consisted of adding

aliquots of gamma emitting isotopes of titanium, scandium,

iron, nickel, yttrium, niobium, thallium and lead to small

volumes of distilled water. The pH of these solutions was

adjusted to various values with dilute HC1 or NaOH. The

solutions were then extracted for one minute with an equal

volume of tetracycline in benzyl alcohol (4.5x10 M) . The

mixture was centrifuged to guarantee complete separation

of the layers. The layers were drawn off and the individual

phases counted with a sodium iodide detector connected to a

scaler and automatic sample changer. Curves of percent extraction into the organic phase versus pH were plotted.

* See disclaimer in Preface.

117
The percent extracted in the diagrams are for one extraction

with tetracycline in benzyl alcohol. Because of the hydrolysis

of the tetracycline, if more than one extraction is desired,

it may be necessary to readjust the pH of the aqueous phase.

For this reason extraction studies will be conducted for

some metals using buffered solutions.

3 . Results

Figure 49 shows the extraction curves obtained for the

metals investigated in the preliminary study. Thallium (III)

extracted much more readily than the other elements but this

extraction can not be attributed to the formation of a thallium

tetracycline complex since thallium extracted equally well

into the solvent, benzyl alcohol. The extraction of the

other elements shown in figure 49 are attributable to the

formation of tetracycline complexes which were soluble in

benzyl alcohol since they did not extract into benzyl alcohol.

Those ions with charges of three or greater extract more

readily than lead or nickel.

/a

100

80

60

40

20

1 1 1 1 i i i i l

Ti

—

\JL \1±L)

sz^f—

III y^

— \AA JPb

1 1

In

i

—

— //// /l\li

\ —

— //// Fe(n) _ — /11 . -

i /i I 1

i 2 3 4 5 6

P H

7 8 10

Figure 49. Extraction curves with tetracycline

118
The difference in the extraction of thallium and iron in

different oxidation states is illustrated in figure 50. The

higher oxidation states of thallium and iron extract much

more efficiently.

Figure 50. Extraction of iron and thallium at various

oxidation states

Since some of the metals which extract in the lower pH

range could not be separated from one another by a pH adjustment, the possibility of differential complexation through the

addition of a complexing agent was investigated. The extraction of yttrium, scandium, iron and niobium were studied with

EDTA, citric acid, and tartaric acid added. The concentrations

of these complexing agents were varied from . 01M to 1M.

EDTA proved to be such a strong complexing agent that none of

these metals were extracted at any pH value.

Figure 51 shows the extraction curves obtained in the

presence of . 01M citric acid. Higher citric acid concentrations did not significantly change the shape or position of

119
100 -

Q 80

LJ

h-

( ) < 60

or

h-X

LU 40

Figure 51. Extraction with citric acid added.

the curves. At pH 9-5 scandium, yttrium, and iron (III) can

be separated from niobium by the addition of citric acid but

no separation of scandium, yttrium, and iron was attained.

Figure 52 illustrates the effect of tartaric acid as a

complexing agent but again no separation of scandium, yttrium,

and iron (III) could be realized.

Although.it was not possible, using tetracycline, to isolate each of the elements studied, it was decided the results

offered enough promise especially for the possibility of group

separations, for the investigation with other elements to

begin. These additional investigations are currently being

conducted by studying the extraction of the transition elements

The data obtained thus far are summarized in figure 53.

120
100 i

1 1 1 I

1

i i.-y

., . • •

-y.

80 — ,:- S SC - Q

LU — /'/ /

u 60

/.' /

<

// /Fe

i— „~ I; /

x 40

LlI J^— _ ^ ySs\

20 ;/..•'/

— ^^ . '

*

' / —

i ^^r>--.V-^" i i i i i 4 5

PH

7 8

Figure 52. Extraction with tartaric acid added

Figure 53(a) illustrates the extraction of the group lb elements (copper, silver, and gold). The extraction of Au(III)

is due to the formation of an ion-associated complex with the

benzyl alcohol. The partial extraction of silver, copper and

gold were attributed to formation of tetracycline complex.

Interesting separations obtained within groups are mercury

from zinc and cadmium, yttrium and scandium from lanthanum,

niobium from tantalum, and molybdenum and tungsten from

chromium.

4 . Conclusions

On the basis of experimental work conducted thus far it

would appear that tetracycline may be applicable to many

chemical separations, both individual and group. The goal of

this project is the study of the extraction of most of the

elements in the Periodic Table with tetracycline. The use

of buffered solutions will be investigated in some instances

as will the effect of various different acids.

121
(a)

100

67

UJ 33

h-o

<

\-

X

UJ

I I I I Aiinm i i Hq

l l l Zn.Cd i i i. i Group I b Group II b 5 100

UJo

ffi 67

o_

33 -

La i i i

1 1 1 1 1 I I L

Zr/

1 I l i i i i I 2 4 6 8

pH

Group IH b

10 2 4 6

pH

Group W. b

8 10

(b)

100 1 1 1 1 1

I I l Nb~~1

67 -

a

UJ 33

o

<

rr

- /

i //i i

i i V

i i Ta i

li

fiwio

/w

i i i iCr i i i i X

UJ Group ¥ b Group 21 b

pH

Group YIH b

Figure 53. Extraction of transition elements with

tetracycline

(L. W. Masters)

122
D. A

T. E

W. D

F. A

Paul

G. W

B. A

7. PERSONNEL AND ACTIVITIES

A . Personnel Listing

Activation Analysis Section

Philip D. LaFleur, Section Chief (1)

Betty E. Crowther, Secretary (1)

Wanda Collins (Summer Worker, Youth Opportunity

Program)

Activation Analysis, Reactor

P. D. LaFleur, Project Leader (1)

Becker, Project Leader (1)

Gills (1)

Kinard (1)

Lundgren (1)

Quittner, Guest Worker, Hungary

Smith (on leave to IAEA) (1/4)

Thompson (1/2)

Kenzie DeLaine (Summer Worker, Youth Opportunity

Program)

Activation Analysis, Cockcroft-Walton Generator

S. S. Nargolwalla, Project Leader and Assistant

Section Chief (1)

J. E. Suddueth (1)

Jerzy Niewodniczanski , Guest Worker, Poland

E. P. Przybylowicz , Industrial Research Associate (1/6)

Sheryl Birkhead (Summer Worker)

Activation Analysis, LINAC

G. J. Lutz, Project Leader (1)

L. W. Masters (1/2)

Radiochemical Separations

B. A. Thompson, Project Leader (1/2)

I. H. Qureshi, NBS Contract, Pakistan (1/3)

W. F. Marlow, Guest Worker - USAEC (2/3)

L. W. Masters (1/2)

Activation Analysis of Biological Materials

D. A. Becker, Project Leader (1/2)

NOTE: Fractions in parentheses indicate fraction of the year

in this position.

123
B . Publications

*R. E. Jervis, H. Al-Shahristani and S. S. Nargolwalla,

"Fast Neutron Continuous Activation Analysis of Dilute Solutions," Proceedings of the 3rd International Conference,

Modern Trends in Activation Analysis, National Bueau of

Standards, Gaithersburg, Maryland, 1968, pp.

S. S. Nargolwalla, E. P. Przybylowicz , J. E. Suddueth

and S. L. Birkhead, "Blank Considerations in 14-MeV Neutron

Activation Analysis for Trace Oxygen," ibid, pp.

E. P. Przybylowicz, Gilbert W. Smith, J. E. Suddueth and

S. S. Nargolwalla, "Activation Analysis of Chloride and

Iodide in Photographic Emulsions Using 14.7- and 2.8-MeV

Neutrons," ibid, pp.

S. S. Nargolwalla, E. P. Przybylowicz, J. E. Suddueth

and S. L. Birkhead, "Solution of Blank Problems in 14-MeV

Neutron Activation Analysis for Trace Oxygen," Anal Chem., hi , 168 (1969).

E. P. Przybylowicz, Gilbert W. Smith, J. E. Suddueth and

S. S. Nargolwalla, "Activation Analysis of Halogens in Photographic Emulsions Using a Neutron Generator," Anal. Chem.,

41, 819 (1969).

S. S. Nargolwalla and P. D. LaFleur, "Characterization of

Standard Reference Materials for Industry and Research by

Neutron Generator Activation Analysis," Proceedings of American

Nuclear Society, Topical Meeting, Radiation and Isotope Technology in Latin American Development, San Juan, Puerto Rico,

1969, pp

L. W. Masters, A. B. Braun, A. Schwebel, "Determination

of Strontium-90 in Environmental Samples After Extraction of

Yttrium-90 and Barium-l40 Using Tetracycline," Anal. Chem.

(in press) .

L. W. Masters, "The Separation and Determination of

Strontium-90 in Environmental Samples Using Tetracycline,"

Master of Science thesis project submitted to the American

University, Washington, D. C., June, 1968.

G. J. Lutz, Determination of Rare Earths by Photon Activatioi

Analysis," Talanta, (in press)

* Work not done at NBS

.

124
B. A. Thompson and P. D. LaFleur, "Rapid Group Radiochemical Separations for Activation Analysis in Steels,"

Anal. Chera. , 4l, 852 (1969).

B. A. Thompson, "Group Radiochemical Separations: A

Practical Approach to Activation Analysis," Proceedings , 3rd International Conference on Modern Trends in Activation

Analysis, National Bureau of Standards, Gaithersburg, Maryland

pp 634-8 (1969)

.

G. J. Lutz, R. J. Boreni, R. S. Maddock, and W. W. Meinke

"Activation Analysis: A Bibliography," NBS Tech. Note 467

(1968).

G. J. Lutz, "Self-Shielding Corrections in Photon Activation Analysis," Proceedings , 3rd International Conference,

Modern Trends in Activation Analysis, National Bureau of Standards, Gaithersburg, Maryland, pp 829-34 (1969).

G. J. Lutz, "The National Bureau of Standards Automated

Activation Analysis Information Retrieval System," Proceedings

,

3rd International Conference on Modern Trends in Activation

Analysis, National Bureau of Standards, Gaithersburg, Maryland,

pp. 1128-30 (1969)

125
C. List of Talks

D. A. Becker, "Neutron Activation Analysis of Ultrapure

Reagents: The Quest for an Absolute Blank," 156 Meeting

of the American Chemical Society, Atlantic City, New Jersey,

September, 1968.

W. D. Kinard, "The Determination of Indium and Copper in

High Purity Gold by Neutron Activation Analysis," (P. D.

LaFleur, W. D. Kinard and D. A. Becker) 156th Meeting of the

American Chemical Society, Atlantic City, New Jersey,

September, 1968

T. E. Gills, "Neutron Activation Analysis of Impurities

in Specially Purified Urea," (D. A. Becker and T. E. Gills)

156th Meeting of the American Chemical Society, Atlantic City,

New Jersey, September 1968.

W. D. Kinard, "The Determination of Indium and Copper in

High Purity Gold by Neutron Activation Analysis," (P. D.

LaFleur, W. D. Kinard and D. A. Becker) 156th Meeting of the

American Chemical Society, Atlantic City, New Jersey,

September, 1968.

G. J. Lutz, "Determination of Yttrium in Rare Earths by

Photon Activation Analysis," (G. J. Lutz and P. D. LaFleur)

156th Meeting of the American Chemical Society, Atlantic City

New Jersey, September, 1968.

G. J. Lutz, "The National Bureau of Standards Activation

Analysis Literature File," 156the Meeting of the American

Chemical Society, Atlantic City, New Jersey, September, 1968.

L. T. McClendon, "Radiochemical Separations Utilizing

Hydrogen bis ( 2-ethylhexyl) Orthophosphate , " (L. T. McClendon,

I. H. Qureshi and P. D. LaFleur) 156th Meeting of the American

Chemical Society, Atlantic City, New Jersey, Spetember, 1968.

B. A. Thompson, "Activation Analysis for Molybdenum in

the Presence of Large Amounts of Tungsten," (B. A. Thompson,

and P. D. LaFleur) 156th Meeting of the American Chemical

Society, Atlantic City, New Jersey, September, 1968.

B. A. Thompson, "Determination of Impurities in High

Purity Zinc by Neutron Activation Analysis," (B. A. Thompson,

and D. A. Becker) 156th Meeting of the American Chemical

Society, Atlantic City, New Jersey, September, 1968.

G. J. Lutz, "Self Shielding Corrections in Photon Activation Analysis," 3rd International Conference: Modern

Trends in Activation Analysis, National Bureau of Standards

Gaithersburg, Maryland, October 7-11, 1968.

126
G. J. Lutz, "The National Bureau of Standards Automated

Activation Analysis Information Retrieval System," (G. J.

Lutz, R. J. Boreni, R. S. Maddock and W. W. Meinke) 3rd

Interantional Conference: Modern Trends in Activation

Analysis, National Bureau of Standards, Gaithersburg, Maryland, October 7-11, 1968.

S. S. Nargolwalla, "Fast Neutron Continuous Analysis of

Dilute Solutions," (J. E. Jervis, H. Al-Shahristani and

S. S. Nargolwalla) 3rd International Conference: Modern

Trends in Activation Analysis, National Bureau of Standards,

Gaithersburg, Maryland, October 7-11 , 1968.

S. S. Nargolwalla, "Blank Considerations in 14-MeV Neutron

Activation Analysis for Trace Oxygen," (S. S. Nargolwalla,

E. P. Przybylowicz , J. E. Suddueth and S. L. Birkhead) 3rd

International Conference: Modern Trends in Activation Analysis, National Bureau of Standards, Gaithersburg, Maryland,

October 7-11, 1968.

E. P. Przybylowicz, "Activation Analysis in Chloride and

Iodide in Photographic Emulsions Using 14.7- and 2 . 8-MeV

Neutrons," (E. P. Przybylowicz, Gilbert W. Smith, J. E.

Suddueth and S. S. Nargolwalla) 3rd International Conference:

Modern Trends in Activaiton Analysis, National Bureau of

Standards, Gaithersburg, Maryland, October 7-11, 1968.

I. H. Qureshi, "Radiochemical Separations for Activation

Analysis Using bis ( 2-ethylhexyl) Orthophosphoric Acid,"

(I. H. Qureshi, L. T. McClendon, and P. D. LaFleur) 3rd

International Conference: Modern Trends in Activation Analysis,

National Bureau of Standards, Gaithersburg, Maryland, October

7-11, 1968.

L. W. Masters, "The Application of the Antibiotic, Tetracycline to Chemical Separations," 4th Middle Atlantic Regional

Meeting, American Chemical Society, Washington, D. C.,

February 12-15, 1969.

P. D. LaFleur, "Activation Analysis in Biology and

Medicine," American Chemical Society Local Section Speaker's

Tour

:

a. Northeast Tennessee Section, Kingsport , Tennessee,

March 10, 1969.

b. Chattanooga Section, Chattanooga, Tennessee, March 11,

1969.

c. Nashville Section, Nashville, Tennessee, March 12,

1969.

127
d. Wilson Dam Section, Muscle Shoals, Alabama, March 13,

1969.

e. Memphis Section, Memphis, Tennessee, March 14,

1969.

f. Auburn Section, Tuskeegee Institute, Alabama,

March 13, 1969

g. Atlanta Section, Atlanta, Georgia, March 18 , 1969.

h. Birmingham Section, Birmingham, Alabama, March 19,

1969.

P. D. LaPleur, "Characterization of Standard Reference

Materials for Industry and Research by Neutron Generator

Activation Analysis," (S. S. Nargolwalla and P. D. LaFleur)

American Nuclear Society, Topical Meeting, Radiation and Isotope Technology in Latin American Development, San Juan,

Puerto Rico, May 4-6, 1969.

S. S. Nargolwalla, "Sources and Evaluation of Systematic

Errors in Neutron Generator Activation Analysis," International

Atomic Energy Agency, Panel Meeting on the Utilization of

Neutron Generators in Physics and Chemistry, Budapest, Hungary,

May 5-9, 1969.

S. S. Nargolwalla, "Neutron Activation Analysis of Gases

in Metals," Center for Professional Advancement Workshop on

"Gases in Inclusions in Solids," Hopatong, New Jersey, June

11, 1969.

128
D . News Releases

"Improved Accuracy in Neutron Activation Analysis," NBS

Technical News Bulletin, 53_, No. 1. January, 1969.

"Rapid Determination of Oxygen in Steel," Steel Times,

March, 1969.

"Neutron Activation Analysis," Chemical Processing, Vol

XV, No. 5, May 1969.

129
E. References

[1] LaFleur, P. D., Editor, NBS Technical Note 45_8 (1968).

[2] Sheely, W. P., NBS Report 9081 (NBSR-10) The National

Bureau of Standards Reactor: Its Description and a Guide to Its Use by Experimentors (1966).

[3] For a discussion of neutron energy determinations and

cadmium ratios, see Becker, D. A. and Smith, G. W.,

NBS Tech. Note 276, (1966).

[4] Thompson, B. A., and LaFleur, P. D., Anal. Chem. 4l,

852 (1969).

[5] Qureshi, I., McClendon, L. T., and LaFleur, P. D.,

Radiochimica Acta, in press.

[6] Qureshi, I., McClendon, L. T., and LaFleur, P. D.,

Proceedings, 1968 International Conference on Modern

Trends in Activation Analysis, in press

[7] Christopher, D. H., and Menis , 0., to be published.

Presented in part at the 4th Middle Atlantic Regional

Meeting, ACS, Washington, D. C, February 1969.

[8] Brune, D., Aktiebolaget Atomenergi Report AE-172 (1965)

[9] NBS Tech. Note 428, (1967).

[10] Brune, D., Acta Chemica Scandenavia, 20_, 1200 (1966).

[11] Brune, D. and Landstrom, 0., Sonderdruck aus Radiochimica Acta, Band 5 (1966).

[12] Girardi , F., and Sibbioni, E., J. Radioanal. Chem. 1,

169 (1968).

[13] Covell, D. P., Anal. Chem. 31, 1785 (1959).

[14] SRM Preliminary Certification, No. 1800, Ion Exchange

Beads, Calcium Loaded, U. S. Dept . of Commerce, Nat ' 1

.

Bureau of Standards.

[15] Lederer, C. M. , Hollander, J. M. , and Perlman, I.

Table of Isotopes (Sixth Ed.) Wiley (1967).

[16] Reynolds, S. A. and Mullins, W. T., International

Journal of Appl. Rad. and Isotopes 1_4, 421 (1963)

[17] Gorsuch, T. T., Analyst 84, 135 (1959).

130
Przybylowicz, E. P., Smith, G. W. , Suddueth, J. E.,

and Nargolwalla, S. S., Analytical Chemistry, in press

Nargolwalla, S. S., Crambes, M. R. , and DeVoe, J. R.,

Anal. Chem. 40, 666 (1968).

Carpenter, B. S., and Cheek, C., Anal. Chem. in press.

Fisher, C, and Cole, M. , The Microscope 1_6, 8l (1968)

Fleischer, R. L., and Price, P. B., Geochim. et

Cosmochim. Acta 28_, 170-4 (1964).

Fleischer, R. L., and Price, P. B., J. Appl. Phys

.

34, 2903 (1963).

Fleischer, R. L., and Price, P. B., Science 140 , 1221 (1963).

Sandell, E. B., Colorimetric Determination of Traces

of Metals , Intersciences Publisher, Inc . , N. Y~. (1959)

Lutz, G. J., and DeSoete, D. A., Anal. Chem. 40, 820

(1968).

Leeson, L., Krauger, J. E., and Nash, R., Tetrahedron

Letters, No. 18, 1155 (1963).

Schumacher, G. E., Amer. J. Hosp. Pharm. 20, 580

(1963).

Albert, A., and Reese, C. W. Nature 177 , 433 (1956).

Albert, A., Nature 172 5 201 (1953).

Maxwell, D. C, Smith, P. J. A., and Wilford, S. P.

Nature 198, 577 (1963)

[32] Richards, V., Lowenstein, J., Phillips, J., and

Armitage, C, Proc. of the Soc. for Experimental Biol,

and Medicine 107, 550 (1961).

[33] Ogawa, E., Fukuda, R., Suzuki, S., and Shibata, K.,

Gunma J. Med. Science 10, 117 (1961).

[34] Ihsen, K. H. , and Urist, M. R., Proc. of the Soc. for

Experimental Biol, and Medicine 109 , 799 (1962).

[35] Ishidate, M. , and Sakaguchi, T., Pharmaceutical

Bulletin (Tokyo) 3, 147 (1955).

131

[18]

[19]

[20]

[21]

[22]

[23]

[24]

[25]

[26]

[27]

[28]

[29]

[30]

[31]
[36] Baker, W. A., Jr., and Brown, P. M. , J. Am. Chem. Soc.

8_8, 1314 (1966).

[37] Masters, L.W., Master of Science Thesis project submitted to the American University, Washington, D. C.,

June 1968 entitled, "The Separation and Determination

of Strontium-90 in Environmental Samples Using

Tetracycline .

132
8 . ACKNOWLEDGMENTS

The Activation Analysis Section wishes to thank Dr.

Robert S. Carter, Mr. Tawfik M. Raby, Mr. James F. Torrence

and the operating crews of the NBS Reactor for making the

irradiation of samples possible. We also wish to acknowledge

the assistance of Mr. Hugh E. DeSpain and the NBS Health

Physics staff who have helped us with the safe handling of

these samples. Further the assistance of Dr. J. E. Leiss

and staff of the LINAC Radiation Division for arrangements

to use the LINAC facilities is gratefully acknowledged. We

also wish to thank Dr. Brian Joiner for assistance in the

statistical analysis of experimental data.

We wish to thank the U. S. Atomic Energy Commission for

making possible the visit of Mr. W. F. Marlow. We wish also

to thank the International Atomic Energy Agency and the

Academy of Mining and Metallurgy, Crakow, Poland, for making

possible the visit of Dr. Niewodniczanski and the Ford Foundation and the Central Institute for Physics, Budapest,

Hungary, for making possible the visit of Dr. Quittner.

This section acknowledges Mrs. R. S. Maddock for her

assistance in preparing this report.

Special appreciation is expressed to Mrs. B. Crowther who

typed the entire report and whose tireless efforts as section

secretary has made our entire operation more efficient and

effective

.

133

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