Nattonal Bureau of Standards

Library, E-01 Admin. Bidg.

JAN 5 B71 NBS TECHNICAL NOTE 548

Summary of Activities

July 1969 to June 1970

IAL BUREAU OF STANSJWSS

/O

UNITED STATES DEPARTMENT OF COMMERCE

Maurice H. Stans, Secretary

NATIONAL BUREAU OF STANDARDS • Lewis M. Branscomb, Director

NBS TECHNICAL NOTE 548

ISSUED DECEMBER 1970

Nat, Bur. Stand. (U.S.), Tech. Note 548, 164 pages (Dec. 1970) CODEN: NBTNA

Activation Analysis Section:

Summary of Activities, July 1969 to June 1970

Philip D. LaFleur and Donald A. Becker, Editors

Activation Analysis Section

Analytical Chemistry Division

Institute for Materials Research

National Bureau of Standards

Washington, D.C. 20234

*»fAU Of

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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 60 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 all 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.

In 1970 we plan to issue these summaries for all of our

sections. The following is the fourth annual report on 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

reasonable accuracy when carefully done.

The Activation Analysis Section is composed of several

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 a project oriented to the analysis

of biological materials; a nuclear data handling project and

project for the utilization of the nuclear track technique

v
for quantitative analysis. Considerable interaction among

the groups is encouraged so that activities of common interest are worked upon jointly in the several groups.

A roster of the personnel in the Section is given in

Part 7. In the past the Section has enjoyed the presence

of scientists from abroad, and from within the United States;

they 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 OF CONTENTS

Page

1. INTRODUCTION 2

2. LIAISON 4

3. ACTIVATION ANALYSIS WITH THE NUCLEAR REACTOR. . . 5

A. Introduction 5

B. Facilities 6

1. Characterization of the NBSR for Neutron

Activation Analysis 6

2. Installation and Operation of a ComptonSuppressed Gamma Scintillation Spectrometer 23

C. Research Activities 28

1. Characterization of Medieval Glass by

Neutron Activation Analysis 28

2. The Determination of Sulfur Using Secondary Reactor Tritons 37

3. Analysis for Iodine in Protein Solutions

by Neutron Activation Analysis 4l

4. Lead Alloy Flux Monitors for Neutron

Activation Analysis 42

5. Extraction of Metals with Bis(2-ethylhexyl) Orthophosphoric Acid 44

6. Determination of Short-lived Trace

Impurities in Sodium Metal by Neutron

Activation Analysis 48

D. Analysis of Standard Reference Materials. . . 50

1. Biological Standard Reference Materials:

Orchard Leaves 50

2. Determination of Cobalt, Silver, and Gold

in SRM Glass by Neutron Activation

Analysis 56

3- Survey of SRM 20 ppb Trace Elements

in Glass 60

4. Survey of SRM 1 ppm Trace Elements in Glass 6l

5. The Determination of Manganese in

Trace Elements in Glass Samples 62

VII
TABLE OF CONTENTS (cont)

Page

6. Intermediate Purity Sulfur 63

7. Potassium Chloride Single Crystal .... 65

8. Determination of Indium and Copper in SRM

High Purity Gold V/ire by Neutron Activation Analysis 66

9. The Determination of Cobalt in CobaltAluminum Flux Wires 69

10. Analysis of Cast Iron Steels for

Tellurium 71

11. Activation Analysis of SRM Steels Using

Group Radiochemical Separations 72

12. The Determination of Cobalt in SRM Steels

by Neutron Activation Analysis 7

h

13. Determination of Copper and Cobalt in SRM

Steel 1265 B-l 75

14

.

Nondestructive Determination of Manganese

and Vanadium in Iron and Steel Standard

Reference Materials 76

15. The Multielement Analysis of SRM Steels

1261 and 361 by Destructive Neutron

Activation Analysis 78

E. Special Analyses 79

1. Determination of Copper Homogeneity in

Copper-Lead Alloy 79

2. Analysis of Neodymium Laser Glass by

Neutron Activation Analysis 80

3. Determination of Trace Elements in Ruby

Laser Crystals 82

4. Determination of Trace Elements in

Synthetic Sapphire Crystals 87

5. Determination of Gold in Semiconductor

Silicon by Neutron Activation Analysis . . 87

F. The Nuclear Track Technique 88

1. Introduction 88

2. Analysis of Uranium and Boron in

Trace Elements in Glass 89

3. Analysis of Orchard Leaves for Boron and

Uranium 94

4. Photo-fission Studies 9h

vm
TABLE OF CONTENTS (cont)

Page

ACTIVATION ANALYSIS WITH THE COCKCROFT-WALTON

NEUTRON GENERATOR

A. Introduction 96

B. Research Activities 97

1. Gamma-ray Spectra and Experimental Sensitivities for 3-MeV Neutron Activation

Analysis 97

2. D-D Neutron Activation Interferences in

14-MeV Neutron Activation Analysis .... 99

3. Thermalization Characteristics of 3-MeV

Neutrons 100

4. Pulse Pile-up and Nuclear Interferences

in the Analysis for Trace Oxygen. .... 103

C. Analysis of Standard Reference Materials

1. Cadmium Cyclohexane Butyrate SRM 1053a. . 105

2. Magnesium Cyclohexane Butyrate SRM 106lc. 105

D. Special Analyses

1. Oxygen in Metallic Barium 106

2. Oxygen in Niobium Metal and NiobiumTitanium Alloy 106

E. Status of Neutron Generator Facility

1. Reconditioning the Penning Ion Source . . 107

2. Health Physics 107

ACTIVATION ANALYSIS WITH THE NBS LINAC

A. Facilities 109

1. Target System 109

2. Sample Terminal Ill

3. Beam Dump Ill

4. Pneumatic Transfer and Shuttle Systems. . Ill

5. LINAC Interface System 113

6. Pneumatic Tube Coupled Lazy Susan .... 114

B. Analyses

1. Determination of Thallium in Glass and

Zinc Using Photon Activation 114

ix
TABLE OF CONTENTS (cont)

Page

2. Determination of Gold, Tantalum,

Manganese and Iron in Cadmium Metal by

Photon Activation 120

3. Determination of Nitrogen in Metals by

Photon Activation 122

6. APPLICATION OF COMPUTERS TO ACTIVATION ANALYSIS

A. Introduction 124

B. ALSPIS - Activation Analysis Peak Area

Program 128

C. A Mini-Computer and the Activation Analysis

Section 135

D. Miscellaneous 139

7. PERSONNEL AND ACTIVITIES

A. Personnel Listing 140

B. Publications l4l

C. List of Talks 142

8. ACKNOWLEDGMENTS 145

9. REFERENCES 146

x
LIST OF FIGURES

Figure Page

1. Positions of the pneumatic tube terminals

in the NBSR 7

Neutron cross sections of copper, gold and

cobalt as a function of neutron energy. . 8

3. Diagramatic representation of the circumferential flux distribution in the NBSR pneumatic

tubes at three positions inside the rabbit. . . n

4. Copper foil placement for circumferential flux

measurements in the NBSR pneumatic tubes .... 12

5. Axial flux distribution in the NBSR pneumatic

tubes 12

6. Axial flux dropoff Inside the rabbits (normalized to proximal end =1.00 for each pneumatic

t ub e ) ]_ 3

7. Device used to measure pressures generated from

irradiated samples 18

8. Vertical distribution of the total neutron flux

In the NBSR LTIF 21

9. Partially shielded Compton-suppressed gammaray scintillation spectrometer with associated

electronics 24

10. Schematic of the Compton-suppressed gamma-gamma

scintillation spectrometer 25

11. Gamma sum spectra of cobalt-60 26

12. Gamma-gamma coincidence spectra of cobalt-60. . 27

13. Gamma-gamma coincidence spectra of cesium-137 . 28

14 . Fragment of Medieval glass with sample cut from

one edge 33

15. French 14th Century circle of stained glass - Angel of the Passion 36

16. Correlation between weight and neutron selfabsorption of lithium fluoride 39

xi
LIST OF FIGURES (cont)

Figure Page

17. Lead alloy ingot being rolled into a thin foil. 43

18. Extraction curves for bis ( 2-ethylhexyl

)

orthophosphoric acid as a function of

hydrochloric acid concentration 45

19. Extraction curves for bis ( 2-ethylhexyl

)

orthophosphoric acid as a function of

perchloric acid concentration 46

20. Extraction curves for bis ( 2-ethylhexyl)

orthophosphoric acid as a function of nitric

acid concentration 47

21. Freeze drying apparatus for the removal of

moisture from biological samples 51

22. Ruby crystal as cut for analysis 83

23. Microscope display for fission track counting . 90

24. Boron distribution across the diameter of a glass wafer 92

25. Photon activation study depicting uranium

weight fraction to total track response .... 95

26. 3-MeV neutron buildup as a function of time at

constant beam current 100

27. Relative flux determinations as a function of

moderator thickness 102

28. LINAC irradiation assembly 110

29. LINAC target assembly 110

30. LINAC beam dump 112

31. LINAC pneumatic transfer system control .... 112

32. LINAC interface system control 113

33- Varian 622/i minicomputer 136

Xll
LIST OF TABLES

Table Page

1. NBSR irradiation facilities 6

2. Experimental neutron flux - NBSR pneumatic

facilities 9

3. Circumferential flux distribution in NBSR

pneumatic tube facilities 11

4. Cadmium ratios for the NBSR pneumatic tube

facility 14

5. Radiation damage and temperature measurements

at 10 MW in RT-1 15

6. Radiation damage and temperature measurements

at 10 MW in RT-2 - 15

7. Radiation damage and temperature measurements

at 10 MW in RT-3 16

8. Radiation damage and temperature measurements

at 10 MW in RT-4. . . . 16

9. Heat resistance of polyethylene rabbits: effect

of irradiation time 17

10. Pressure measurements on neutron irradiated

samples 19

11. Cadmium ratios in the LTIF 22

12. Circumferential flux variation in the NBSR LTIF

irradiation facility 22

13. Results of activation analysis of simulated

ancient glasses 31

14. Homogeneity of Medieval and 18th Century

glasses 32

15- Results of analysis of Bourges glasses 33

16Alkali ratios found in Medieval glasses .... 34

17. Manganese/sodium ratios found in Medieval

glasses 35

Xlll
LIST OF TABLES (cont)

Table Page

18. Results of activation analysis of Medieval

window

19. Results of sulfur analysis

20. NBS biological Standard Reference Materials . .

21. Moisture removal for orchard leaf samples . . .

22. Manganese homogeneity: orchard leaves

23. Manganese homogeneity: orchard leaves

2k. Combined averages for all containers

25. Results of analysis of 1 ppm glass (SRM 615). .

26. Results of analysis of 50 ppm glass (SRM 613) .

27. Results of analysis of 500 ppm glass (SRM 611).

28. Glass impurities

29. Analytical results for 1 ppm glass

30. Manganese concentrations in trace elements in

glass samples

31. Impurities found in intermediate purity sulfur.

32. Variation in impurities found - surface vs.

interior of sample

33. Analyses for copper and indium in SRM gold wire

34. Analysis results for cobalt in cobalt-aluminum

wire 70

35- Results of tellurium analysis in SRM cast iron

steel 72

36. Results of analysis of SRMs 1263, 1264, 1265. . 73

37- Results of analysis of SRM ductile irons. ... 73

38. Results of analysis for cobalt 75

xiv

37

ho

51

52

54

55

56

58

59

59

60

61

63

64

65

69
LIST OF TABLES (cont)

Table Page

39. Results of activation analysis of SRM 1265 B-l. 76

40. Results of manganese and vanadium analyses of

iron and steel Standard Reference Materials . . 77

41. Results of multielement analysis of SRM steels

1261 and 36l 78

42. Results of activation analysis of neodymium

glasses 82

43. Results of activation analysis of ruby

crystals, group I 84

44. Results of activation analysis of ruby

crystals, group II 84

45. Results of activation analysis of ruby

crystals, group III 85

46. Detection limits for trace elements in ruby . . 86

47. Trace elements observed in sapphire crystals. . 87

48. Results of analysis of silicon wafers 88

49. Boron and uranium content in the trace elements

in glass (50 ppm) 91

50. Boron and uranium content in the trace elements

in glass (1 ppm) 93

51. Boron and uranium content in the trace elements

in glass (0.02) 93

52. Experimental sensitivity for 3-MeV neutron

activation analysis 98

53. Results of trace oxygen interferences 104

54. Results of analyses for thallium in NBS glass . 117

55. Thallium content in zinc and zinc salts .... 118

56. Representation of alphanumeric character string

internally and as base ten numbers 131

xv

ACTIVATION ANALYSIS: SUMMARY OF ACTIVITIES

July 1969 to June 1970

Edited by

Philip D. LaFleur and Donald A. Becker

This survey of the activities of the Activation Analysis

Section covers the period July 1969 to June 1970. An overview

cf the progress made by the various projects in the Section

is given. A discussion of the operating characteristics of

the NBS Reactor for activation analysis is given and the

research activities of the various projects are discussed.

Special attention is given to the many practical problems

encountered by the Section during the year, and the results

of the solutions to these problems, especially the analysis

of Trace Elements in Glass Standard Reference Material (SRM)

and of the botanical SRM. The quantitative analysis for

boron using the nuclear track technique is discussed as well

as extension of the anlaysis for uranium using this technique.

An extensive discussion of the value of 3-MeV neutrons

in neutron generator activation analysis and the experimental

parameters for using these neutrons is discussed. The new

target and pneumatic transfer systems at the NBS LINAC for

photon activation analysis are described and the application

of photon activation analysis for determining thallium In

glass and high purity metals and gold, manganese and iron In

high purity cadmium is discussed. Research into determination

o.f nitrogen by photon activation analysis is described.

A computer program, ALSPIS, for peak finding and integration is described and the application of a mini-computer

to the Activation Analysis Section is noted.

Key words: activation analysis; Cockcroft-Walton neutron

generator; computer programming; NBS LINAC ^ NBS Reactor;

Standard Reference Materials

.
1. INTRODUCTION

During the past year the Activation Analysis Section has

reached a turning point. For the past several years we have

been engaged in demonstrating our capabilities for analysis

both within and outside of the National Bureau of Standards.

We now feel that we have satisfactorily demonstrated this

capability and intend during the coming year to spend considerably more of our time on research activities and the

development of new analytical techniques.

As was mentioned in last year's progress report, our

most serious problem at that time was the attainment of a

capability using computers for processing data. The addition

of Dr. H. P. Yule to our staff has started us on the solution

to this problem and we are looking forward, in the next

several years, to the fruitful application of automated data

processing both on large digital computers and on mini

computers

.

During this year the NBS Reactor has operated all year

at a power level of 10 Megawatts. The operating schedule

has been approximately three to four weeks at full power and

then about three days of down time for routine maintenance.

The flight lines in the four originally installed pneumatic

tube systems have been replaced with armoured polyethylene

tubing and all four systems are operative and are being fully

characterized. The pneumatic tube system in the thermal

column facility should be installed very early in fiscal

year 1971 as should the 7/8 in. pneumatic tube system.

A comprehensive study of the application of 3-MeV neutrons

from neutron generators and of the problems of 3-MeV neutron

buildup in 14-MeV activation analysis and of the use of

moderated 3-MeV neutrons has been completed. This was the

final phase of the research with the neutron generator laboratory and the application of this machine in the near future

will be primarily in the analysis of samples.
During this year a new bremsstrahlung converter and rabbit

system has been installed in the NBS LINAC . This has caused

the facility to be down during the last four months of the

year, however, the new system should be operable in August of

1970. Work has been completed on the analysis of carbon

in a number of high purity metals using the LINAC, and efforts

have begun on the analysis of elements other than the light

elements in a number of samole matrices.
LIAISON

'

During this year we have enjoyed a fruitful cooperative

program with the Smithsonian Institution and Mrs. Jacqueline

Olin of that organization. This program has resulted in some

very interesting data on the characterization of stained glass,

so that future work may be extended for the elucidation of

the origin and/or period of manufacture of stained glass

windows

Fledgling cooperative programs have also begun with the

Internal Revenue Service of the Treasury Department, in this

case with Mr. C. Michael Hoffman and Mr. Maynard Pro of that

laboratory, in the analysis of bullet lead. We are expecting

to become more deeply involved in forensic science and are

looking forward to an interesting and fruitful collaboration

with the Treasury Department

With the occasion of the closing down of the Naval Research Laboratory Reactor, we have been participating with

the Reactor Radiation Division with accomodating new users

to the NBS Reactor and expect to have many encounters with

these new users during the next year. We anticipate additional

cooperative programs may result from these contacts. It is

our opinion that the cross-fertilization resulting from

activities of this kind is beneficial to both groups who

participate, and we welcome inquiries about the use of NBS

facilities either on a cooperative basis, or where the interested party merely wishes to utilize the radiation source.
3. ACTIVATION ANALYSIS WITH THE NUCLEAR REACTOR

A . Introduction

During this year we have continued our efforts to attain

a high degree of versatility in activation analysis and have

spent much of our time in developing a greater expertise in

nondestructive activation analysis, particularly coupled with

computer data reduction. We intend, of course, to continue

our efforts in obtaining the highest possible versatility in

both destructive and nondestructive activation analysis and

in the investigation of parameters affecting accuracy and

precision in activation analysis.

We have spent a great deal of time in characterizing the

two pneumatic tube facilities which became available this

year. All four of the originally installed facilities are

now available and have been very carefully studied in terms

of temperature effects, relative neutron fluxes and to a more

limited extent on neutron energy spectra. We have maintained

a very satisfactory relationship with the Reactor Operations

Section and with the Health Physics Section.

A greater portion of our activities this year than in

past years has been spent in service analyses for organizations

outside the Bureau of Standards and for other groups within

the Bureau of Standards. Particularly notable are our efforts

in the analysis of lasers and silicon chips. We are presently

beginning a service analysis program for the Department of

Agriculture, involving the determination of mercury in wheat

and peas and with the National Institutes of Health in the

determination of iodine in blood serum samples. It is our

hope that contacts with other governmental organizations in

the Washington area, such as these, may continue and that

interesting cooperative research programs may develop therefrom.
B. Facilities

1 Characterization of the NBSR for Neutron Activation

Analysis

Since the last report [1] much additional information

has been obtained on the irradiation characteristics of the

NBSR. This information has included thermal neutron flux

variations within the irradiation areas, neutron energy

distribution in the various irradiation positions, temperature effects, and measurements of the gas production from

several different biological materials. These different

types of measurements will be treated separately here, with

the techniques used presented along with the results obtained.

For comparative purposes, all results obtained to date will

be included here.

The current status of the NBSR irradiation facilities

are given in table 1. The operational facilities include

Table 1. NBSR irradiation facilities.

Pneumatic Facilities

Ident

.

Position

Access

Attitude Diameter Status

RT-1 Core gap edge

RT-2 In core

RT-3 Core gap, D„0

RT-4 DpO reflector

RT-V Vertical thimble

RT-C } Vertical thimble

(RT-V-2)) (Converter) J-4

RT-T Thermal column

Manual Access Facilities

Vertical

thimble

Vertical

thimble

In core, G-2

Reflector, V-4

Horiz. 1.5 in. Operational

Horiz. 1.5 in. Operational

Horiz. 1.5 in. Operational

Horiz. 1.5 in. Operational

Vert. 0.75 in. Under const.

Vert. 0.75 in. Under develop

Vert. 0.75 in. Under const.

Vert. 1.5 in. Under const.

Vert. 1.5 in. Operational

Vert. 1 in. Planning

(7 Positions)
the four pneumatic tubes (fig. 1) and the manually accessed

Long Term Irradiation Facility (LTIF) located in the G-2 v vertical thimble. The program of characterization has

included measurements on all of these operational facilities,

but with particular emphasis on the pneumatic facilities

which are used for the majority of the activation analysis

irradiations

.

a. Pneumatic Tube Irradiation Facilities

The four operational pneumatic tube facilities have

diameters of 1 1/2 inches, are powered by C0 ?

, and terminate

in significantly different neutron environments (fig. 1).

TANK WALL

RT-1 W////,W//,

////////

RT- 2 '///////W//,

Figure 1. Positions of the pneumatic tube terminals

in the NBSR. (Terminology: The proximal

end of the rabbit is that end closest to

the reactor core vertical centerline; the

d istal end is that furthest from the core

centerline .

)
Linear polyethylene rabbits, which have a capacity of approximately h2 cm ( 2 . 6 cm dia x 8 cm long) are used in these facilities [1].

( 1) Neutron Flux Measurements

Initial measurements in the pneumatic tubes were for the

thermal neutron flux. Of interest was the absolute thermal neutron flux, and the variation of this neutron flux within each

irradiation position. Some measurements also have been made

of the neutron energy distribution characteristics within

each facility.

( 2

)

Thermal (Sub-cadmium) Neutron Flux

The thermal (sub-cadmium) neutron flux measurements were

made using the thin foil technique [2]. Two different

detector foils, gold and copper, were used for these measurements. Copper foils were preferred, due to the absence of

significant resonance peaks in its neutron cross section

(fig. 2). Some measurements were also made on gold foils,

COPPER

GOLD

COBALT

UT

(B)

NEUTRON ENERGY (EV)

Figure 2. Neutron cross sections of copper, gold and

cobalt as a function of neutron energy.
since gold is often used in other laboratories. However,

since gold has significant neutron self-absorption problems,

slight variations in foil thickness can lead to appreciable

errors in flux determinations

The technique used consisted in irradiating a thin foil

(0.005 in Cu foil; 0.001 in Au foil) of the element of

interest in the appropriate facility, then counting the

radioactive species formed on a large volume Ge(Li) detector.

This detector has been calibrated for absolute photopeak

counting efficiency using NBS calibrated radioactive sources.

Neutron fluxes were then calculated from nuclear constants.

The thermal (sub-cadmium) neutron flux was obtained after

correction for epicadmium contribution to the foil activity.

For copper, this correction did not exceed 7%, however, for

gold this correction was in excess of 25%. No neutron selfshielding corrections were made.

The results of the flux determinations are found in

table 2 . It is apparent that the thermal neutron flux

in tubes RT-1, RT-2 and RT-3 are very similar, with a significantly lower flux in RT-4. The gold foils gave higher

i

, flux numbers (relative to those from Cu measurements) for

the tubes RT-1, RT-2 and RT-3, where a larger epicadmium

neutron flux exists, but a value very close to that for

1 copper in RT-4, where the neutrons are highly thermalized.

|| Repeated determinations of these flux measurements gave

results agreeing well with the original values.

Table 2. Experimental neutron flux-NBSR pneumatic facilities

Sub-Cadmium Neutron Flux-Proximal End

l "2 -In

( n. cm sec ) Detector

Foil RT-1 RT-2 RT-3 RT-4

Gold 7.0xl0 13 6.9xl0 13 5-8xl0 13 1.4xl0 13

Copper 5.6xl0 13 6.3xl0 13 4.9xl0 13 1.3xl0 13
( 3) Neutron Flux Variation Measurements

In order to obtain accurate and precise results in

neutron activation analysis using a nuclear reactor, the

geometrical variation of the flux within each irradiation

position must be accurately evaluated. Previous studies of

the NRL Nuclear Reactor [3] revealed serious neutron flux

variations within the sample irradiation area, and thus emphasized the need for such measurements.

The flux measurements for the NBSR were made on the

inside dimensions of the polyethylene rabbits for the 1 1/2

in. pneumatic tube system (RT-1 to RT-M). Since the irradiation space available for each facility approximated a right

circular cylinder 2.6 cm in diameter and 8 cm long, both

the circumferential flux distributions" and the axial flux

distributions were measured. The procedure developed used

3/16 In. disks of copper foil or 20 mil copper wire with gross

gamma counting of the copper-64 radioactivity using an automatic sample changer. The copper activity was checked for

extraneous radioactivities by counting on a 47cc Ge(Li) detector and 4096 channel PHA and by following the decay characteristics of the foil on the gross gamma counter for over 8

half-lives. Both procedures revealed no detectable radioactivity in either the wire or the foil other than copper-64.

All samples from a particular type of measurement were counted

under identical conditions.

The results obtained for the circumferential flux distribution of the pneumatic facilities are found in table 3-

In all cases copper foils were used. These results are also

shown graphically in figure 3- The foil placement diagram for the circumferential flux measurements is shown in

figure 4. It can be readily seen from these results that

there is no significant circumferential flux variation at

any plane perpendicular to the axis of the right circular

cylinder comprising the rabbit irradiation area.

10
Table 3. Circumferential flux distribution in

NBSR pneumatic tube facilities.

Pneumatic

Pacili

Tube

ty

Location in

Rabbit

N

Mo

o. of

nitors

7

7

7

Relative

Neutron Flux

RT-1

RT-1

RT-1

Proximal

Middle

Distal

1.000±0.003b

0.9^7+0.004

0.871+0.007

RT-2

RT-2

RT-2

Proximal

Middle

Distal

7

6

7

1.000±0.005

0.902±0.002

0.874±0.003

RT-3

RT-3

RT-3

Proximal

Middle

Distal

7

6

7

1.000±0.004

0.928±0.003

0.829+0.003

RT-4

RT-4

RT-4

Proximal

Middle

Distal

7

7

6

1.000±0.010

0.787+0.007

0.603+0. 014

,Normalized to proximal end = 1.000.

Uncertainties are ts//n at 95% confidence level

POSITION RT-1 RT-2 RT-3 RT-4

Figure 3. Diagramatic representation of the circumferential flux distribution in the NBSR pneumatic

tubes at three positions inside the rabbit.

11
Figure 4. Copper foil placement for circumferential flux

measurements in the NBSR pneumatic tubes.

The axial flux measurements were made using the copper

wire. A single length of the wire was centered along the

rabbit axis and held securely in position with polyethylene

foam. After irradiation, the wire was cut into 1 cm sections,

weighed, and counted. The results, shown in figures 5 and

6, reveal a serious neutron flux dropoff along the axis

of the rabbit. The magnitude of this dropoff over the

length of the rabbit ranged from approximately 11% for RT-1

1.00

X

.80

Z

* .60

D

UJZ

uj .40

>

<

-I£ .20 .RT-4

I 2 3 4 5 6 7 8

DISTANCE FROM PROXIMAL END (CM)

Figure 5. Axial flux distribution in the NBSR pneumatic tubes.

12
1.00

.90

.80

.70

.60

2 3 4 5 6

DISTANCE FROM PROXIMAL END (CM)

Figure 6. Axial flux dropoff inside the rabbits (normalized to proximal end =1.00 for each pneumatic

tube )

.

to over k0% for RT-4. The flux dropoff was linear in all

pneumatic facilities except RT-2.

( 4 ) Neutron Energy Distribution Measurements

Some preliminary neutron energy distribution measurements have been made, consisting almost entirely of cadmium

ratio measurements . These cadmium ratios were made at both

ends of all four pneumatic facilities. Three different

elemental foils were used, 1 mil thick gold foil, 5 mil copper

foil, and 2 mil cobalt foil. Interlocking 20 mil cadmium

covers were used for the cadmium shielded measurements.

The results of these measurements are found in table

4. The values for gold and cobalt were obtained primarily

13
Pneumatic Location

Facility In Rabbit

RT-1 Proximal

RT-1 Distal

RT-2 Proximal

RT-2 Distal

RT-3 Proximal

RT-3 Distal

RT-4 Proximal

RT-4 Distal

Table 4. Cadmium ratios for the NBSR

pneumatic tube facility.

pi Experimental Cadmium Ratios

Au(Cd) C u(Cd) C o(Cd)

7.3 46 29

10 64

3.7 15.7 11

4.0 15.8

10.3 65 42

16.5 80.5

87 536 395

111 582

a For elemental foils using 20 mil cadmium covers.

for intercomparisons with other laboratories, since these two

elements are the ones generally used for flux and cadmium

ratio measuarements . However, as mentioned previously, the

large resonance neutron cross sections (fig. 2) casts some

doubt on usefulness of measurements using these foils.

The copper foils, which behave as true "infinitely thin"

foils throughout the neutron energy range seen in a nuclear

reactor, should yield results which can be compared directly

to neutron fluxes above and below the cadmium cutoff.

Additional efforts will be expended in the direction of

neutron energy measurements in these facilities by the use

of threshold detectors. These threshold detectors yield

integrated neutron fluxes above their reaction thresholds,

and will be used to plot diagrams of the integral neutron

flux vs. neutron energy similar to the work done by us at

the NRL nuclear reactor [3]«

( 5 ) Temperature Measurements

An extensive series of measurements were made in the

pneumatic facilities using stick-on type temperature sensors.

14
These sensors were described previously [1] as were the

various techniques employed and some preliminary results

.

Essentially all of the measurements have been completed at

this time, and the results obtained are presented in tables

5, 6, 7 and 8. The equilibrium temperatures were lower than

those expected, and therefore maximum irradiation times are

based on factors other than temperature, e.g. radiation

Table 5. Radiation damage and temperature

measurements at 10 MW in RT-1

.

Material and Configuration

Empty rabbit

Quartz ampoule, empty

Lead in quartz ampoule

Iron in quartz ampoule

Length of

Wt.(g) Irrad

.

6 h

a Results

Radiation

damage not

excessive^

3.• 5g 20 m 170±10°F

0.

3.

.6g Pb

. 5g Quartz 20 m 225±5°F

0,

3,

. 6g Pe

. 5g Quartz 20 m <180°F

,20 minutes required to reach equilibrium temperature

In this and succeeding tables, this phrase means the rabbit

is not so brittle as to break upon deceleration in receiver

Table 6. Radiation damage and temperature

measurements at 10 MW in RT-2.

Length of

Materials and Configuration Wt.(g) Irrad. Results

Empty rabbit

Empty 2 dram poly vial

20 m

3.25g poly 10 m

Empty heat sealed sections 0.7g poly

of poly tubing (total) 10 m

155±5°P

205±5°P

205±5°F

15
Table 7. Radiation damage and temperature

measurements at 10 MW in RT-3-

Material and Configuration

Empty rabbit

Empty rabbit

Lead pieces (6) in sealed

quartz ampoule

Quartz block

Quartz block

Lead and quartz - lead

(single piece) in sealed

quartz ampoule

Empty rabbit, containing

two 2/5 dram poly vials,

with polyvial packing

Length of

Wt.(g) I rrad

.

20 m

Results

— <110°F

- 8 h Radiation

damage not

excessive.

0.6g Pb

3.7g Quartz 4 h 195±5°F

6.0g 15 m 170±10°F

7.0g 20 m 190±10°F

l.Og Pb

3.2g Quartz 10 m >220°F

8 h Radiation

damage not

excessive to

rabbit or

polyvials

.

Polyfoam

crumbly

.

Table 8. Radiation damage and temperature

measurements at 10 MW in RT-4.

Length of

Material and Configuration Wt . ( g

)

Empty rabbit

Empty rabbit

Lead in sealed quartz

ampoule

0.6 g Pb

3-5 g quartz

Irrad

.

Results

16 h Radiation

damage not

excessive

16 m 105±5°F

15 m <130°F

damage to the sample carrier. Irradiations of up to 8 hours

are made routinely in the polyethylene rabbits at a thermal

13 -2 -1

neutron flux of 5x10 n.cra sec

16
Initially, reactor operations would permit temperatures

of up to 200°F in the rabbit facilities. However, a few

tests were made of the melting point of these rabbits and

the effect of irradiation on the melting point. The results

of these tests are given in table 9- As a result of these

tests, the maximum permissable temperature in the pneumatic

facilities was raised to 240°F.

Table 9. Heat resistance of polyethylene rabbits:

effect of irradiation time a . Length of Effect of Temperature (°p)

Irradiation

none

10 sec

1 min

230°F 240°F 260°F 280°F

none none slight completely

softening melted

none none slight mostly

softening melted

none none slight completely

softening deformed

10 min none none slight transparent

softening and starting

to deform

Melting point of conventional polyethylene is ^250°F; high

b density polyethylene ^270°F.

Rabbits irradiated in RT-1.

(6) Pressure Measurements

When the request was made for the irradiation of freeze

dried biological materials in the NBSR, a number of questions

were raised which could not be definitively answered. One

of these questions involved the danger of explosion of a quartz ampoule due to the pressure generated by gaseous

decomposition products of the material under Irradiation.

To effectively answer this question, and to establish criteria

for the safe irradiation of a variety of aqueous solutions,

17
organic materials, and biological materials, a series of pressure measurements was initiated. The pressure testing system

used to test irradiated samples is shown in figure 7- It

consists of an appropriate sized rubber balloon, a section of

polyethylene tubing which will fit inside the balloon but is

still large enough to hold the irradiated quartz ampoule, and

a gas-tight clamp for the open end of the balloon. To make

a pressure measurement, an irradiated sample (inside the

quartz ampoule) is inserted into the tubing and the balloon,

the end of the balloon clamped shut and the ampoule struck

smartly with a hammer. The polyethylene tubing prevents the

shattered quartz from puncturing the balloon, and the pressure

inside the ampoule estimated from the degree of expansion of

the balloon.

Tests with a hypodermic syringe (less needle) snugly

3 fitted to the balloon opening Indicated that 1/2 cm of gas

could be readily detected. The quartz ampoules being tested

had volumes of 1 cm or greater, therefore excess pressures

of 1 atm could most certainly be measured. Since even poor

quality glass blowing produces ampoules which can readily

withstand 2 atm excess pressure, and good quality quartz

glass blowing can produce ampoules which will withstand

internal pressures in excess of 5 atm excess pressure, a preliminary limit for testing purposes of 1 1/2 atm excess

pressure was established.

RUBBER BALLOON

CLAMP

POLYETHYLENE TUBING

Figure 7- Device used to measure pressures generated from

irradiated samples

18
The results of these pressure tests are found in table

10. Prom these results it can be seen that freeze-dried

biological materials produce only small quantities of gaseous

products upon irradiation in the pneumatic facilities RT-1,

RT-3 and RT-4. For example, 1/2 gram of freeze-dried orchard

leaves, irradiated for 6 hours in RT-3 at a neutron flux of

13 -2 -1 3 5x10 n.cm sec produced only about 1 cm of gaseous decomposition products (which produced approximately 1 atm of

excess pressure in the ampoule size used) . These conditions

of excess pressure are well within the tentative limits set

up for these tests.

Table 10. Pressure measurements on neutron

irradiated samples.

Sample

Length

Wt (g) of Irrad. Facility Results a

Orchard leaves 0.25

freeze dried

Alfalfa, freeze 0.25

dried

Blood, freeze 0.25

dried

Liver, freeze 0.25

dried

Water, in heat- 2.0

sealed polyethylene

Orchard leaves, 0.50

freeze dried

Orchard Leaves, 0.50

freeze dried

2 h

2 h

2 h

2 h

5 m

RT-1 Small pressure

buildup

RT-3 No significant

pressure buildup

RT-3 No significant

pressure buildup

RT-3 No significant

buildup

RT-3 No significant

pressure buildup

8 h RT-4 Small pressure

buildup

6 h RT-3 Small pressure

buildup

a 3 Limit of detection ^1/2 cm ; "small pressure buildup"

estimated to be 1/2 - 1 cnP

.

19
b . Long Term Irradiation Facility (LTIF)

The flux measurements made on the LTIF Vertical Thimble

(see [4] for details of construction) included only neutron

flux measurements and cadmium ratio determinations . Since

this facility has its own recirculating D

? cooling system,

the temperature measurements were not immediately essential.

The measurement of pressures due to sample decomposition have

not yet been made on this facility, at least partly due to

the probable applicability of those measurements made with

the pneumatic tubes.

The flux variation measurements on this facility were

made using two different techniques. For the vertical flux

measurements a copper wire inside of sealed polyethylene

tubing was quickly inserted the entire length of the vertical

irradiation facility and irradiated for 30 m at a power level

of 10 KW . At the end of the irradiation time, the copper

wire was quickly removed from the reactor, the polyethylene

tubing removed, and the copper wire cut into small pieces for

gross y counting of the 12. 9h copper-64 radioactivity. At

the lower end, where the activity level was relatively high,

the wire was cut into 1/2 inch lengths . As the induced radioactivity decreased in the upper portions of the wire (above

the core area), sections were increased to 1 inch, and finally

to 6 inches at a point considerably above the upper core area.

After this vertical flux diagram was obtained, the piece

having the highest specific activity (from the core gap area)

was taken for an absolute flux determination.

Circumferential flux measurements were made during

irradiations for the cadmium ratios. These measurements were

taken using suitably weighted polyethylene rabbits at the end

of a nylon fishing line. The copper wire was wrapped around

the rabbit in which the bare copper foil was irradiated. The

cadmium covered copper foil had its separate irradiation.

This set of two irradiations was necessary for each area

measured. Four areas were measured this way, as shown in

20
figure 8 and table 11.

The results of these measurements are shown in figure

8 and tables 11 and 12. Figure 8 shows the vertical

distribution of the total neutron flux in the LTIF, and

demonstrates the effectiveness of the core gap area as a region of maximum thermal neutron flux. The maximum neutron

;

:

70

NEUTRON FLUX DISTRIBUTION

IN VERTICAL THIMBLE G-2

(J

/

/

—^-<j

UPPER CORE

SI\

-v

CORE GAP

LOWER CORE

>v^

100 90 80 70 60 50 40 30 20 10

PERCENT OF MAXIMUM FLUX

65

60

55

50

45

40

35

HEIGHT

(inches)

30

25

20

15

10

5

[Figure 8. Vertical distribution of the total neutron flux

in the NBSR LTIF.

21
Table 11. Cadmium ratios in the LTIF.

Cadmium Ratio

Area H eight (

0.5

in. ) (Copper)

Below core 84.1

Lower core 11 32.8

Core gap 19.5 54.9

Above core 56 1830

Table 12. Circumferential flux variation in the

NBSR LTIF irradiation facility.

Location in LTIF

Area Height (in)

Location in

Rabbit

No. of

Monitors

Relative . Neutron Flux

On bottom

On bottom

On bottom

0.5

1.8

3.0

Bottom

Middle

Top

9

9

9

1.000±0.010 c 1.122±0.008

1.238±0.009

Lower core

Lower core

Lower core

11.5

12.8

14.0

Bottom

Middle

Top

9

9

9

1.000±0.018

1.055±0.026

1.088±0.015

Core gap

Core gap

Core gap

20.0

21.3

22.5

Bottom

Middle

Top

9

9

9

1.000±0.021

0.997±0.0l6

0.980±O.Ol8

Upper core

Upper core

Upper core

28.5

29.8

31.0

Bottom

Middle

Top

9

9

9

1.000±0.08l

0.837±0.170

0.725±0.092

Above core

Above core

Above core

56.5

57.8

59.0

Bottom

Middle

Top

9

9

9

1.000±0. Oil

0.885±0.013

0.774±0.011

, Measured from bottom of polyethylene rabbit (see figure 8)

Normalized to bottom = 1.00.

c —

Uncertainties are ts//n at 95% confidence level.

22
flux for 10 MW operation was found to be approximately

14 -2 -1 1.1x10 n.cm sec . The cadmium ratios are given in table

11. The core gap cadmium ratio of 55 is significantly

higher than the core area, which has a cadmium ratio of

approximately 33- All of these cadmium ratios are somewhat

higher than would normally be expected, due to the highly

thermalizing effect of the D

? filled thimble. The copper

cadmium ratio of 1830 at a hight of 56 inches in the thimble

12 -2 -1 (with a 10 MW neutron flux of ^5x10 n.cm sec ) suggests

the potential use of this facility for obtaining a highly

thermalized neutron environment. The results of the circumferential flux distributions in the LTIP reveal a small but

significant variation in the neutron flux from side-to-side

in the thimble. This variation is strongly dependent upon

the area chosen for the irradiation, but usually is less

than 10% over the outer diameter of the polyethylene rabbits

used for the measurements.

c . Conclusions

The NBSR has been characterized for many of its irradiation parameters, both for the pneumatic facilities RT-1, RT-2,

RT-3 and RT-4, and for the LTIF (G-2 Vertical Thimble). While

some additional measurements remain to be made, the irradiation facilities available for use are revealed as a set of

extremely useful and convenient facilities under which a wide variety of irradiation environments can be obtained.

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

2 - Installation and Operation of a Compton-Suppressed

Gamma Scintillation Spectrometer

The Compton-suppressed gamma-gamma coincidence scintillation spectrometer, with its associated equipment and lead

shielding was obtained from the Naval Radiological Defense

Laboratory [5] and has been installed and made operational,

(fig. 9). The individual modular equipment has been checked

for damage during shipment and was found to be in proper

23
Figure 9- Partially shielded Compton-suppressed gammaray scintillation spectrometer with associated

electronics

.

order. The system has been interconnected and successfully

operated in the gamma sum, and the gamma-gamma sum coincidence modes . a. Modes of Operation

( 1) Gamma Sum Anticompton

In this mode, the signal pulses from each of the two

central crystals are linearly summed into a preamplifier,

24
then fed to an amplifier. Two outputs are taken from the

amplifier, a delay output directly fed to a 400 channel analyzer as signal, the other, a prompt output in coincidence

with the annulus signal gated to block the signal pulse any

time the sum crystal pulse is in coincidence with the annulus

The block diagram of this mode of operation is shown in

figure 10a.

Pc A C

SCA

A

SCA

C

(a)

COINC.

SIGNAL

BLOCK

400 Ch

ANALYZER

RIDL

34-27

u

— p9

SCA,

SCA.

SCA,

(b)

COINC. PULSE

INVERTER COINC. INPUT

SIGNAL

400Ch

ANALYZER

RIDL

34-27

C,-CRYSTAL I C 2- CRYSTAL 2

A,-ANNULUS I A2-ANNULUS 2

P*-PR0MPT0UTPUT

D- DELAYED OUTPUT

P, - PREAMPLIFIER FOR ANNUL

P2 - PREAMPLIFIER FOR CRYSTAL I P

3 - PREAMPLIFIER FOR CRYSTAL 2

A, - AMPLIFIER FOR ANNUL

A 2 - AMPLIFIER FOR CRYSTAL I A3 - AMPLIFIER FOR CRYSTAL 2

SCA-SINGLE CHANNEL ANALYZER

COINC.-COINCIDENCE UNIT

Figure 10. Schematic of the Compt on-suppressed gamma-gamma

scintillation spectrometer. (a) Block diagram of

the coincidence system used in the gamma sum mode

with compton supression. (b) Block diagram of the

coincidence system used in the gamma-gamma coincidence sum mode with Compton suppression.

25
(2) Gamma- Gamma Sum Coincidence

In this mode, only the pulses from the two central

crystals which are in coincidence are accepted. These pulses

are linearly summed after the amplifiers and fed as a signal

to a 400 channel analyzer. All pulses from the annulus in

coincidence with the signal pulses were fed to a blocking

gate to provide Compton suppression. The block diagram of

this mode of operation is shown In figure 10b.

b. Results

60,

( 1) Gamma Sum Ant i compton

The system was used to count "Co in the gamma sum and

the gamma sum Compton-suppressed modes . The Compton suppression reduced the Compton continuum by about a factor of H while reducing the background a factor of about 2. A Comparison of spectra of Co with and without Compton-suppression

is shown in figure 11.

A. GAMMA SUM

NO COMPTON

SUPPRESSION

B. GAMMA SUM

WITH COMPTON

SUPPRESSION

Figure 11. Gamma sum spectra of cobalt-60

26
( 2) Gamma- Gamma Sum Coincidence

The system was used to count Co and Cs in the gammagamma coincidence mode with and without Compton suppression.

Non-coincidence events were reduced by a factor of between

15 and 20. The best results were obtained using gamma-gamma

coincidence with Compton suppression as can be seen in figures

12 and 13.

A. GAMMA SUM

WITHOUT COMPTON

SUPPRESSION

B. GAMMA-GAMMA SUM

COINCIDENCE WITHOUT

COMPTON SUPPRESSION

C. GAMMA-GAMMA SUM

COINCIDENCE WITH

COMPTON SUPPRESSION

Figure 12. Gamma-gamma coincidence spectra of cobalt-60

27
A. GAMMA SUM

B. GAMMA-GAMMA SUM

COINCIDENCE

Figure 13- Gamma-gamma coincidence spectra of cesium-137.

(H. L. Rook and D. A. Becker)

C . Research Activities

1. Characterization of Medieval Glass by Neutron

Activation Analysis

a. Introduction

nesearcn carried out by J . S. Olin of the Smithsonian

Institution and E. V. Sayre of Brookhaven National Laboratory

[6] and by I. Perlman and F. Asaro of the Lawrence Radiation

Laboratory [7] has shown that neutron activation analysis

can be a valuable aid in determining the provenance of

pottery samples. The Smithsonian was interested in learning

whether similar methods could be applied to the characterization of medieval glass to help decide questions of the

28
authenticity of stained glass windows, which panes are original and which replacements , and possibly even the precise

origin of a partciular window. To determine the feasibility

of this approach, a joint study was begun using the NBS

reactor and personnel from both the Smithsonian and NBS, as

reported previously [8].

b . Experimental Procedure

The first experiments were carried out to gain information

about the composition and homogeneity of medieval glass and

also to find out whether reliable data could be obtained from

very small samples (10-100 mg) since this is probably close

to the maximum which could be expected to be made available

from an actual window. The general approach followed was

similar to that used by Coleman and Wood [9] in their work

with plate glass fragments. First, a small sample OlO mg)

was irradiated for 2 seconds at 6x10 n.cm . sec and used

to obtain information on major constituents such as Na, K,

Ca, Al, Cu, and Sb . Vanadium and indium were also observed

if present in these samples. A larger sample (50-100 mg)

was then irradiated for 15 min-1 hr to obtain information

about minor and trace constituents. Initial plans were to

24 dissolve these large samples and to remove Na with HAP

122 124 and ' Sb by extraction of the chloride complex with

isopropyl ether. However, because of the high levels of

Cu and K present in many of the samples, the value of the

1 HAP separation was reduced and therefore this step was eliminated and the only chemistry used was separation of Sb from

samples containing high concentrations of this element.

A 47-cc Ge(Li) detector was used for analysis of all samples.

To check out the reliability of the procedure, analyses

were carried out on samples of simulated ancient glasses which

had been prepared by Corning and distributed by Dr. Robert Brill

to various interested laboratories, including the Smithsonian,

for the purpose of checking different analytical methods. The

29
results obtained with these glasses are shown in table 13 and

it is clear that even in these very preliminary analyses, the

reproducibility is quite satisfactory.

Because of the high manganese levels in samples "A", "B",

and "D" , the analyzer dead time during the measurement of

Al and V activities was often as high as 30-40%. Although

no gain shift or loss of resolution could be observed, the

high dead time created a problem in the choice of suitable

standards for these elements. The problem was solved by

using NBS Standard Reference Material (SRM) steels, which

often have high Mn concentrations, as standards. SRM 3^ } which has 0.57% Mn and 1.16% Al, was used for an Al standard,

and SRM 30f with 0.79% MN and 0.18% V for a V standard.

Using samples of these, it was fairly easy to get a standard

56 with about the same Mn activity and thus about the same

dead time as the sample. This eliminated the need for decay

corrections on these short-lived activities. SRM lOle with

1.77% Mn was used as a Mn standard.

The results shown in Table 13 were all obtained on

glasses which had been prepared under modern conditions and

should have had good homogeneity. The samples to be used

for medieval glass analysis had been obtained by Mrs . Olin

from the The Cloisters, Metropolitan Meseum of Art, through

the courtesy of Miss Jane Hayward and included two sets of

fragments. One set was blue glass, designated as probably

13th-l4th century, and the other was a greenish "bullseye"

glass designated "Paris - 1775." Although this second set

could not really be considered medieval, these two sets of

fragments were used to determine the degree of homogeneity

to be expected from glasses which had not been prepared

under modern conditions. Some of the results obtained are

shown in table 14 and it can be seen that in cases where

there are plenty of counts and no interferences, e.g. Na

in both glasses and Co and La in the blue glass, good precision

30
Table 13. Results of activation analysis of

simulated, ancient glasses, %.

II A ft llD I! II p It Tl T-\ 1

Na

Cu

Mn

Sb

Ba

Co

K

V

12.34

12.46

12.82

11.64

14.09

14.56

13-93

13.36

0.990

0.951

0.853

0.891

1.197

1.020

1.044

1.11

1.08

1.04

1.05

2.27

2.34

2.24

2.22

1.07

1.01

0.92

0.95

0.312

0.310

0.280

0.816

0.793

0.787

0.190

0.179

0.183

0.0007

0.0011

0.417

0.474

1.50

1.24

1.26

1.12

0.31

0.32

0.30

0.34

- 0.87

0.66

O.67

- -

8.71

8.85

7.68

8.08

-

0.11 0.030

0.034

0.118

0.124 0.014

- - 2.30

2.07

2.39

9.74

9.62

0.0046

0.0042

0.017

0.019

0.018

0.015

0.0395

0.0394

0.0088

0.0093

0.47

0.43

0.44

2.21

2.19

2.22

0.422

0.419

0.419

2.56

2.54

Al

can be obtained. Where there are few counts or interfering

peaks, e.g. K and Fe, the precision is much poorer. On the

basis of these results it was concluded that differences of

less than ±10$ could not be regarded as significant for most

elements

.

31
Table 14. Homogeneity of Medieval and 18th Century glasses,

Concentrations, %

Clear "Bullseye" Glass

Paris, 1775

4.08 ±0.05°

3-51 ±0.62 c 0.0015 ±0.0001 C

0.00066±0. 00006°

0.57 ±0.06°

0.404

1.66

Blue Glass, XIII-XIV

Element Cent. (?)

Na 0.96 ±0.04 a

K 18.9 ±2.7 a

La 0.004o4±0.00008b

Co 0.114 ±0.001b

Fe 0.93 ±0.25b

Cu 0.38 ±0.03 a

Sb 1.08

Al 0.82

, Std. deviation of the mean; n=4

n-5

n-8

Also observed in the blue glass: In, Ca, Sc, Zn, Rb

.

Also observed in the "bullseye" glass: Ca, Sc, Zn, Sm

The data in table 14 also represents the results of a study of various cleaning methods and it was concluded that

the specific type of cleaning is relatively unimportant.

c . Results on Medieval Glass Samples

In addition to the two sets of fragments, five glass

samples, all designated as being from Bourges, XII or XIII

century, were analyzed using the procedure described above.

Figure 14 shows how these glasses were sampled. This is a typical piece, about 2 In. long, with a cut-out portion along

one edge where samples for analysis were removed. Samples

from this glass were used for one of the homogeneity experiments and so more sample was removed than would be the case

for an ordinary analysis. The results obtained on these five

samples are listed in table 15.

32
Figure l4 . Fragment of medieval glass with sample cut from

one edge. (Photograph courtesy of the

Smithsonian Institution.)

Table 15. Results of analysis of Bourges glasses, %.

Element CI ear Glass Blue Glass

No. 1 No. 2 No. 3 No. 4 No. 5

Na 9.8 1.45 4.2 0.45 14.5

K <2 13.8 3.8 10 .0 <2

Mn 0.50 0.57 0.85 0.47 0.36

Al 0.47 0.55 1.20 1.05 1.13

Co 0.0012 0.0008 0.0006 0.0012

Cu 0.24 0.024 0.026 0.067 0.26

Sb - - - - 1.24

La <0.0007 0.0044 0.0015 0.0036 <0.001

Fe 0.16 0.58 0.44 0.35

Ca 9 9 ^11 12 4

Other elements observed in some or all cases: Sc a Zn, Sm.

Rb j and As.

A consideration of the procedures and techniques used

in making medieval glass provided a basis for the selection

of useful comparisons and correlations of the various results

in light of the original objectives. This information was

obtained from the excellent series of papers' on ancient

glass and glassmaking by Turner [10] and the medieval

treatise of Theophilus, "On Divers Arts" [11]. In the

33
Middle Ages glass was made by mixing "ash" and "sand" with

perhaps an additive or two to produce a desired color. It

was not until the 18th century that the differences between

K and Na began to be known and, therefore, the ratio of the

alkalis in medieval (and ancient) glasses would have been

determined entirely by the ratio in the particular "ash"

used. Medieval glass is, in general, characterized by a high K content which results from the use of vegetable ash,

particularly beechwood ash, for glassmaking. Turner points

out that the alkali content of such ash varies not only with

the type of ash but also from place to place and from time

to time depending on soil content, etc. He also points out

that at various times ash from seacoast plants with high Na

was popular.

In view of these considerations it seemed appropriate to

calculate alkali ratios for the various glasses analyzed.

These are listed in Table 16 and it can be seen that the

glasses fall into groups with very different ratios, well

outside the ±10$ limits mentioned above.

Table 16 . Alkali ratios found in Medieval glasses.

Sample % Na % K K/Na

Blue fragments 0.95

Bourges No. 4 0.45

Bourges No . 2 1.45

Bourges No . 3 4.2

"Bullseye" fragments 4.0

Bourges No . 1 9.8

Bourges No . 5 14.5

18.9

10.0

13.8

3.8

3.5

<2

<2

19-9

22.2

9.5

0.90

0.88

<0.2

<0.l4

Turner also mentioned that beechwood ash has an unusually^

high Mn content so that the source of the Mn in the glasses

might again be the "ash" rather than a decolorizing agent or

34
other additive. In this case the Mn-alkali ratios might also

be significant. Table 17 gives these ratios for the same

glasses and division into the same groups can be made, although the agreement is less striking than in the case of

the K/Na ratios

Table 17. Manganese/sodium ratios found in Medieval glasses.

Sample % Na % Mn Mn/Na

Blue fragments 0.95 1.08 1.1*1

! Bourges No. 4 0.45 0.47 1.04

1 Bourges No. 2 1.45 0.57 0.39

'Bourges No. 3 4.2 0.85 0.20

; "Bullseye" fragments 3.9 0.40 0.10

Bourges No. 1 9.8 0.50 0.051

Bourges No. 5 14.5 0.36 0.025

Rb was detected in all three of the high-K glasses

(Bourges no. 2, no. 4 and the blue fragments) and, as can be

seen from tables 14 and 15, glasses analyzed can be divided

into almost identical groups on the basis of the La concentrations. Much more work needs to be done on the treatment

of the data to gain the maximum amount of information, but

before attempting this, data should be obtained on samples

! whose provenance Is known much more accurately than is the

case for the samples available for this study.

d. Analysis of Panes from Medieval Windows

The Gellatly Collection of the National Collection of

Pine Arts, Smithsonian Institution, contains a group of

; stained glass windows whose attribution is French, 14th

Century. Mr. Charles Olin was studying one of these windows

"The Angel of the Passion," which is shown in figure 15.

He observed that although the three blue panes on the right

of the window appeared similar in color under ordinary light,

under infra-red light the lower pane appeared to be different

35
Figure 15 • French 14th Century circle of stained glass

Angel of the Passion. (Photograph courtesy

of the National Collection of Fine Arts,

Smithsonian Institution.)

from the other two. By lifting the leading around the panes

and very carefully chipping the edges he was able to remove

small samples from each of the three panes. Two 10-mg and

one 50-mg samples of each pane were analyzed by the neutron

activation procedure described above and the results are

shown in table 18. It is clear that the middle pane, not

the lower, is very markedly different from the other two.

The upper and lower panes are more like each other than they

are like the middle pane, but they are far from identical.

The difference in the appearance under infra-red light may

possibly be explained by the substantially lower cobalt concentration in the lower pane relative to the other two,

since it is this constituent that produces the blue color.

In any case the results show very clearly the risks in drawing conclusions about an entire window from the analysis of

a single pane or by a single technique.

36
Table 18. Results of activation analysis of Medieval

window, %.

Element Upper Pane Middle Pane Lower Pane

Na . 0.293

0.294 ( 2.40 ( 0.212

2.49 2.50

< 2.56

I H.65

11.33 11.48

1 11.86

0.302 ) 0.300 2.49 \ 2.50 0.207) 0.184

( 0.319 ' 2.56 ( 0.224

K 16.73

j 16.76 ( 11.65 ( 19.92

17.24 j 17.25 11.33 j 11.48 20.49 j 21.21

( 18.22 ( 11.86 / 20.33

Mn , 1.11

x lk , 1.18 , 0.8.6

1.09 \ 1#o6 1.14 \ lmll u. ry \ 0>757

Cu 0.129

0.127

Al 1.48

1.45

Co 0.0705

0.0717

Ca 13

La 0.00266

0.00277

Sb 0.00436

0.00477

K/Na 57

Mn/Na 3 .

0.462

0.445

0.131

0.114

0.517

0.501 0.878

0.0920 0.0441

6 8

0.00412 0.00207

0.0123 0.00230

4.4 104

0.45 3.9

In, Zn and Sc present in all panes, Ta present in lower pane

only

(J. S. Olin, E. C. Miller, and B. A. Thompson)

2 . The Determination of Sulfur Using Secondary Reactor

Triton_s_

The determination of sulfur is of great importance in the

metallurgical, biological, and environmental sciences. The

most widely used method for this determination utilizes bomb

combustion followed by volumetric or gravimetric analysis.

With complete combustion, this technique maintains the accuracy of gravimetric or volumetric analysis, but is time37
consuming and does not lend itself to large numbers of

samples . Tritons of 2.7^ MeV produced in a reactor via the

Li(n,a) H reaction have been used for the determination of

oxygen utilizing the 0(t,n) F reaction [12], However,

the method reported has never been totally successful for

low concentrations of oxygen due to the large oxygen contamination in salts of lithium-6 used as sources of tritons . In

the case of sulfur, this type of interference should be

negligible due to the low sulfur content in isotopically

enriched lithium-6 salts.

The use of the S(t,n) reaction was investigated

[13] and more recently Albert et al. suggested the analytical utility of the reaction because of its specificity

below ^13.5 MeV and the high natural abundance of sulfur-32

[1*0. However, this reaction has not yet been analytically

exploited because of the lack of accelerators with the

authorization to accelerate tritium.

An activation analysis technique using reactor produced secondary tritons to induce the S(t,n) CI reaction

has been investigated for the rapid determination of sulfur

In fuel oils. In this work, the 2.78 MeV tritons were used

as they are slightly above the 2.38 MeV threshold calculated

for the 32 S(t,n) 3,(mCl reaction. The calculated reaction

threshold was made using a non-relativistic charge interaction equation [15]. The useful range of the 2.78 MeV

triton in fuel oil is about 2x10 cm. Thus, a thin film of

oil coating a lithium fluoride crystal will stop the triton.

However, to use solid lithium fluoride coated with oil, the

average particle size of the lithium fluoride must be a

constant

.

a.- Experimental

Approximately 25 mg of solid lithium fluoride was weighed

into a quartz ampoule, two drops of oil allowed to coat the

38
salt, and the ampoule sealed. The samples were irradiated

for 15 seconds in the RT-4 facility of the NBS Reactor at a 13 -1 -2 thermal neutron flux of ^1.3x10 n.sec cm . After irradiation, the samples were washed into a separatory funnel with

4 cc of benzene. The ionic chlorine 34m was extracted with

water containing 25 mg of inactive chloride carrier. The

chloride ion was precipitated as silver chloride, washed and

counted for 2 half-lives of the chlorine-34m. The benzene

phase was washed into a 4 dram vial, brought to 6.0 ml volume

and counted in a 5 in. x 4 in. well Nal (Tl) crystal.

b . Results and Discussion

Residual fuel oils of known sulfur content (SRMs 1621 and

1622) were prepared and analyzed by the above procedure.

Samples of SRM 1622 were analyzed using different weights of

10

12.0 14.0 16.0 18.0 20.0 22.0 24.0

WT. LiF IN mg -» Figure 16 . Correlation between weight and neutron selfabsorption of lithium fluoride.

39

26.0
LiP to obtain a correlation between weight and self absorption of the lithium fluoride (fig. 16). This correlation was

used to correct the countrates of the silver chloride precipitates and these results used for analysis. The results

are given in table 19.

Since the benzene phase results were not consistent, the

lithium fluoride was ground in a ball mill to reduce the

particle size and triturated in a low sulfur content oil.

A suspension of 100 mg of LiF per ml of oil was mixed with

an equal weight of standard sulfur oil and analyzed by the

previously mentioned procedure. The initial experiments

showed no neutron self-absorption if less than 0.25 g of oil

mix was used. Analytical results are currently being obtained.

Table 19. Results of sulfur analysis.

Wt LiF Using A°ppt

Sample A° appt A°Benzene % Eff. in Mg % S

1622-1 186.2 153.6 56.1 12.09 Std.

1622-2 277.7 240.7 53-8 24.00 Std.

1621-1 107.1 120.7 46.1 17.22 0.98

A° appt A°Benzene % Ef

186.2 153.6 56.1

277-7 240.7 53.8

107.1 120.7 46.1

106.2 135.2 44.1

108.0 137.5 44.1

134.2 140.8 48.8

130.5 124.7 51.2

1621-2 106.2 135.2 44.1 14. 16 1.06

1621-3 108.0 137.5 44.1 16.98 1.00

1621-4 134.2 140.8 48.8 24.42 1.05

1621-5 130.5 124.7 51.2 24.24 1.02

Average 1.02±.04

Certified Value 1.05±.02

A°=cpm at end of irradiation

(H. L. Rook

40
3. Analysis for Iodine in Protein Solutions by Neutron

Activation Analysis

In collaboration with scientists of the thyroid group in

the Clinical Endocrinology Branch of the National Institute

of Arthritis and Metabolic Diseases, preliminary investigations were made into the possibility of using neutron activation analysis for the determination of low levels of iodine

The samples of Interest were dialyzed protein solutions,

both normal and iodinated. Expected iodine levels ranged

from 10 ng/ml to 1000 ng/ml (0.01 to 1 ppm)

a. Experimental

Initial tests were made on protein solutions with varying

amounts of added iodide. Reagent grade potassium iodide was

used as the standard, with care taken to prevent oxidation

of the iodide to free iodine during the analysis

A procedure was developed which allowed rapid analysis

of samples (^20 m each, including irradiation and counting)

ranging from 5 ng/ml to 10 mg/ml or higher. All the samples

and standards were encapsulated in specially cleaned, heat

sealed medical grade polyethylene tubing. The procedure

developed included irradiation in RT-3 of the NBSR for 5-0 m

13 -2 -1 (thermal neutron flux ^5x10 n.cm sec ), removal from the

polyethylene rabbit, washing the outside of the polyethylene

container in dilute HnO q (1:1), rinsing several times in

distilled H„0, and then drying, positioning in the counting

holder, and counting for 10.0m with a 60cm Ge(Li) detector

and 1024 channel PHA . The counting geometry was varied from

1.25 cm to 20 cm from the detector to compensate for differences in radioactivity levels between samples. All geometry

positions were calibrated with Iodine-128 standards to determine efficiency factors. Flux normalization of the various

irradiations were made with copper foil flux monitors.

b

Results and Discussion

The results of analysis for the preliminary tests yielded

the expected straight line relationship between various

iodine concentrations and the 0.441 MeV radioactivity. The

41
lack of interference using this gamma ray peak was verified

by half-life determinations. In addition, the half-life

determination verified that no significant amount of iodine

escaped by diffusion through the polyethylene during this

period. Considering that free iodine easily diffuses

through polyethylene, and the oxidizing nature of ionizing

radiation, this question was a real concern.

Blank levels were determined for the polyethylene tubing

and for the dialyzing solution, and were found to be approximately 5ng/0.5ml, respectively. Actual samples received

varied from l8ng/0.5ml to 17 . 7yg/0 . lml

.

The results obtained using neutron activation analysis

agreed well with the values obtained from wet chemical techniques for the high iodine level solutions. The low level

iodine samples could not be checked this way, since for

these levels the wet chemical technique was not reproducible.

(D. A. Becker and W. D. Kinard)

4 . Lead Alloy Flux Monitors for Neutron Activation

Analysis

Since the NBSR came into routine use for neutron activation analysis, the high neutron flux available has encouraged the development of a variety of materials as flux

monitors for long irradiations in high flux positions . In

particular, alloys of lead with a number of different elements is in progress

.

The requirements for flux monitors for high precision

and accuracy in activation analysis has been discussed elsewhere [16, 3]. Examination of commonly available elements

for their neutron activation and decay characteristics as

well as their metallurgical properties reveal several that

could be conveniently used. Lead was chosen as the dilutant

due to Its lack of gamma ray producing radioactive products

after neutron irradiation and because of its density and

malleability. Dilute alloys of lead with many elements

42
should remain ductile to allow rolling into foil form. After

production, these foils must then be examined for their suitability as flux monitors.

The elements which have been alloyed with lead to produce foils are copper, gold, zinc, and indium. Selenium was

attempted but has not yet been alloyed before combustion

occurs. At present, only the copper-lead alloy and the goldlead alloy (0.1% element in lead) have been tested for

homogeneity. Both of these elements seem to behave quite

well, yielding with care a homogeneous melt which can then

be rolled into ^5 mil foil (fig. 17) and cut into small

(^5 nig) pieces for use as flux monitors. While reagent

grade lead was used for testing purposes, only 99 -999% purity

material was used for the final products. The homogeneity

of actual 0.1% copper-lead alloy flux produced in this laboratory averaged ±1.1% S , , about twice that of pure copper

Figure 17- Lead alloy ingot being rolled into a thin foil

43
flux monitor foils [31- The homogeneity of gold-lead alloy

foils was slightly higher, at ^±2%.

Efforts have been made to interest a private company in

producing the copper-lead alloy on a large scale for flux

monitor use. However, several attempts by this company

failed to produce a foil with acceptable specifications.

Additional attempts are presently being made either to find

a different alloy which can be mass produced, or to find

another method of producing an acceptable copper-lead alloy.

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

5. The Extraction of Metals With Bis ( 2-ethylhexyl ) Qrthophosphoric Acid

The solvent extraction studies using bis ( 2-ethylhexyl)

orthophosphoric acid (HDEHP) as the extractant which were

described previously [17] have been largely concluded. The

extraction of 53 elements was studied from 1 to 11M solutions of hydrochloric, perchloric and nitric acids into

0.75M HDEHP in petroleum ether. The extraction curves of

these elements are shown in figures 18, 19, and 20. At the

present time no more data has been acquired as to the mechanism of extraction using this reagent, but we are planning

to investigate this mechanism in the future. At the present

time effort is going into the application of these extractions

to group and individual element separations for use with

activation analysis, an area which we believe will be most

fruitful.

(I. H. Qureshi and P. D. LaFleur)

44
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6 . The Determination of Short-lived Trace Impurities in

Sodium Metal by Neutron Activation Analysis

The determination of trace elements in sodium has recently

received much interest because of the use of sodium in liquid

metal cooled reactor technology. For liquid sodium reactors

the determination of corrosion products demands state of the

art analytical techniques.

To measure the extent of corrosion in the coolant lines

of liquid-metal-cooled fast reactors, it is necessary to

measure the build-up of the elements from the stainless steel

piping present in the coolant as a function of operating

time. Because of the high chemical reactivity of sodium

metal, it is very difficult to maintain the integrity of the

sample from the time it is drawn until it is analyzed.

Activation analysis has been largely ignored as a technique for analysis of a sodium matrix because of the intense

activity of the sodium-24 created during irradiation. The

analysis of elements producing short-lived (<15 hrs ) isotopes

is especially difficult. The high energy gamma rays and its

long half-life makes nondestructive analysis impossible.

Destructive analy-sis requires a number of time consuming

radiochemical separations. The time required to allow the

matrix to decay is especially critical for the determination

of elements such as Mn, Cu, and W, etc. since all have halflives of 24 hrs or less. In addition, waiting a week or more

to begin processing the samples after irradiation is usually

undesirable. Because of the radiation hazards involved with

handling the samples, involved chemical manipulations are

very unattractive.

Recently Girardi and Sabbioni [18] reported the use of

hydrated antimony pentoxide (HAP) to remove radiosodium from

biological samples. Decontamination factors of _>10 were

reported for sodium removal from 12m HC1 solutions. The

only other element of the 60 elements they studied retained

48
by the HAP was Ta, which was also quantitatively removed.

The activation analysis of glass samples, containing lk%

NapO, indicated that HAP was indeed as effective for radiosodium removal as was indicated by Girardi and Sabbioni.

Based on these experiments HAP was used to remove the matrix

activity to allow the analysis of short-lived trace impurities in sodium metal

Experimental

Samples of sodium weighing ^100 to 250 mg were encapsulated in medical grade polyethylene or quartz tubing and

irradiated for 12 minutes to 2 hrs at a flux of ^5x10

-2 -1

n.cm sec in the pneumatic tube facilities of the NBSR.

After irradiation, the sample was lightly etched in ethanol

and then dissolved in a 60% ethanol-40% water mixture. The

sample was evaporated to dryness with a heat lamp, carefully

taken up with concentrated HC1, and diluted to 25 ml in a volumetric flask. A 1.0 ml aliquot is taken for subsequent

determination of the sodium content and therefore the total

amount of sodium taken for analysis. The remaining solution

is passed through a 1-cm diameter column containing ^5g of

HAP. After the entire solution has passed through the HAP,

the column is washed with 10 to 15 ml of 12M HC1, and the

wash combined with the main sample . The sample was then

counted using a 47cc Ge(Li) detector in conjunction with a 4096 channel analyzer. In the spectrum energy range 0-3.5

m tt n^ •-, 64^ 56,, 122 cv 87 n 1 , 42^ MeV, one could easily see Cu, Mn, Sb , Br, and K.

As an example of the decontamination achieved with HAP, only

24

a small peak from Na could be seen in the eluate from a column having a radiation level of 4R/hr. The manipulations

involved are readily adaptable to remote handling.

To see if these elements represented the bulk of the

matrix, several etches were made and it was found that the

impurities observed were representative of the entire sample.

(T. E. Gills and P. D. LaFleur)

4 9
D. Analysis of Standard Reference Materials

1 . Biological Standard Reference Materials: Orchard Leaves

After several years of looking into the need for and preparation of a number of biological Standard Reference Materials

[19] an NBS Standard Reference Material, Orchard Leaves, is

now in the process of being analyzed and certified (table 20).

At present, six different botanical materials have been acquired (orchard leaves, tomato leaves, citrus leaves, alfalfa,

pine needles and aspen chips) of which one, orchard leaves,

has been tested for homogeneity, radiation sterilized, packaged,

and is undergoing analysis. This one botanical material will

be certified for major, minor and trace elemental content.

In addition to these botanical materials, efforts are being

made to obtain large samples of lyophilized beef liver and

lyophilized bovine serum. These materials will also be prepared and analyzed for elemental content, and issued as

Standard Reference Materials

The total number of elements in these Biological SRM's

which will finally be certified is still uncertain. At present,;

the elements most likely to be certified include nitrogen,

phosphorus, potassium, calcium, magnesium, sodium, boron, iron,

manganese, aluminum, zinc, copper, arsenic and iodine.

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

a. Determination of Moisture Content in SRM Orchard Leaves

In preparation for the analysis and certification of this

material, a literature search revealed no standardized technique for drying a biological sample before analysis . Therefore , several of the different suggested techniques were

tested in order that one or two effective drying methods

could be suggested for this materials. The methods used included freeze drying (temperature not to exceed 50°C) (fig. 21),

vacuum oven drying at 90°C, oven (air) drying at 90° C, and

oven (air) drying at 135°C.

50
Table 20 . NBS biological Standard Reference Materials.

Material Acquired Packaged Homogeneity Certified

Botanical

Orchard leaves Yes Yes Yes In process

Citris leaves Yes Yes In process

Tomatoe leaves Yes No - Alfalfa Yes Yes In process

Pine needles Yes No - Aspen chips Yes No -

Blood

Bovine serum

Tissue

Beef liver

In process

In process

Figure 21. Freeze drying apparatus for the removal of

moisture from biological samples

.

51
Table 21. Moisture removal for orchard leaf samples.

Drying Drying Visual Weight Average

Technique Time Temperature (°C) Change Loss (% ) Wt . Loss(%)

Freeze

Drying

24h <50 None 3.39

3. 40

3.40

Vacuum oven 24h 90 None 3.05

3.59

3-32

(Air) oven 24h 90 None 3.44

3.53

3.48

(Air) oven 24h 135 Yes 7.60

7.29

7.45

Dessicator

[Mg(C10

4

) 2]

96h 20°C None 1.87

1.96

2.05

1.96

^AOAC methods.

Material turned light brown.

The results of these tests are given in table 21. Sample

sizes were approximately 22 grams each. They were stored in

dessicators for cooling prior to weighing. The results show

that the first three drying techniques used, freeze drying,

drying in a vacuum at 90°C, and drying in an air oven at 90°C,

are all approximately equally effective without apparent loss

of volatile materials. However, drying in an oven at 135°C

obviously results in the loss of some of the sample, and is

not recommended. Dessicator drying is not recommended due

to the length of time required for complete drying. With

only the dessicator, over half of the moisture is removed in

four days. However, the remaining moisture would most likely

require much longer to remove.

A few additional measurements were made on the rate of

moisture pickup in an open laboratory room. These measurements revealed that in 20m almost 0.3% moisture was regained.

In 45m standing, 0.7% moisture was regained, or almost one

quarter of the total moisture lost on drying. The foregoing

52
results emphasize that these samples or any dried samples

should not be exposed to the open air for long without redrying the samples

.

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

b . Homogeneity of Botanical Standard Reference Material : Orchard Leaves

The first of these materials to be studied was the

sample of dried powdered orchard leaves . Before proceeding

with the analyses it was necessary to show that the distribution of elements in the 220 kg of sample was homogeneous

.

Other Sections within the Analytical Chemistry Division

also participated in this study. The Activation Analysis

Section studied the distribution of Mn in the samples

.

The sampling technique used was designed by Dr. J. Mandell,

Statistical Consultant to the Institute for Materials

Research, to satisfactorily sample the 10 barrels containing

the sample

.

( 1) Experimental

The samples as received from OSRM were in tightly closed

bottles. A portion of each sample was transferred to a clean plastic tray and freeze-dried over night. After

removal from the freeze dryer, the samples were stored in

several freshly cleaned dissicators over calcium sulfate.

Individual samples of ^250 mg were accurately weighed and

prepared for irradiation. SRM Steel No. lOle, containing

1.77% manganese, was used as the standard. The samples

were irradiated for 10 seconds In the NBSR RT-4 pneumatic

tube facility. Counting was delayed for ^2.5 hours to

27 allow Mg (t, /? =9.5 min) to decay. Samples and standards

were counted with Identical geometries on a large volume

Ge(Li) detector coupled to a 4096 channel PHA . After

corrections for decay and flux normalization, the manganese

concentrations were calculated.

53
(2) Results and Discussion

Results of the individual analyses are shown in tables

22 and 23. The average of all the samples is shown in

table 24.

Table 22. Manganese homogeneity: orchard leaves.

Sample No. Sample Position

0-2 P - top

0-2 Middle center

0-2 Bottom center

0-2 Q - middle

0-2 Right bottom

0-2 Top center

0-2 Overall average

0-3 Top center

0-3 Right bottom

0-3 Bottom center

0-3 Q - middle

0-3 P - top

0-3 Middle center

0-3 Overall average

0-5 Top center

0-5 Right bottom

0-5 Bottom center

0-5 Q - middle

0-5 P - top

0-5 Middle center

0-5 Overall average

Manganese Cone,

(ppm) Average

104. , 103 103.

98.4, 102. 100.

94.2, 99.8 97.0

100. , 101. 100.

101. , 100. 100.

100. , 106. 103.

101. ±1 .9

101. , 99.6 100.

107- , 104. 106.

102. , 98.6 100.

103-, 96.3 99.6

103-, 97.4 100.

108. , 99.6 104.

102. ±2 .8

99.5, 105. 102

101. , 107 104

98.7, 105. 102.

101. , 99.0 100.

103-, 99.2 101.

98.6, 97.4 98.0

101. ±2 .06

ts//n @ 95% confidence level.

54
Table 23. Manganese homogeneity: orchard leaves.

Sample No. Sample Position

Middle center

Manganese Cone,

(ppm) Average

0-6 102. , 95.4 98.7

0-6 Q - middle 97.0, 95.7 96.4

0-6 Top center 96.3, 97.4 96.8

0-6 Bottom center 93.9, 99.7 96.8

0-6 Right bottom 96.7, 110. 103.

0^6 P - top 100. , 106. 103.

0-6 Overall average 99,1±3.3

0-9 P - top 101. , 96.4 98.7

0-9 Center top 122. b

, 99.6 99.6

0-9 Bottom center 97-4, 97.6 97.5

0-9 Middle center 97-8, 93.0 95.4

0-9 Q - Middle 95.8, 96.0 95.9

0-9 Right bottom 101. , 94.1 97.6

0-9 Overall average 97 .2±1.7

0-10 Center top 99.1, 99.0 99.0

0-10 P - top 97-9, 97.1 97.5

0-10 Center bottom 97.4, 95-3 96.4

0-10 Center middle 98.9, 96.9 97.9

0-10 Q - middle 97-7, 94.0 95.8

0-10 Right bottom 98.1, 94.8 96.4

0-10 Overall average 97..2±1.2

^ts//n at 95% confidence level.

This value rejected in subsequent calculations

55
Table 24. Combined averages for all containers.

Sample No. Average Manganese Cone, (ppm)

0-2 101.

0-3 102.

0-5 101.

0-6 99.1

0-9 97.2

0-10 97.2

Overall average (all samples) 99.6±2.17

Range 93.0 - 108

ts//n at 95% confidence level.

In general, the results indicated that, for manganese,

these samples are homogeneous for 250 mg to at least ±3% • All results were calculated on a dry weight basis

.

(W. D. Kinard)

2. Determination of Cobalt, Silver, and Gold in SRM

Glass by Neutron Activation Analysis

The determination of Co, Ag and Au was required for

certification of the SRM trace elements in glass. In many

samples Co and Au can be determined nondestructively and Ag

can be determined using a simple AgCl separation. However,

in the case of the SRM glasses many uncommon elements were

added and all the trace elements were added at about the

same concentration levels. This resulted in different interferences than are ordinarily encountered and, thus, a different approach to the analysis

.

The radionuclides produced by neutron activation of Co,

Ag and Au all have half-lives long enough to permit decay of

24 the very high Na levels before analysis, and so the HAP

separation for Na used in previous work [8] was not necessary.

In this work the glass disks (0.2-0.8 g) were irradiated for

13 —2 -1

1 hour at a flux of ^6x10 n . cm .sec and set aside for

56
7-10 days. After this time the principal radioactive species

*. i a ^o 198. I82 m 152^ l4o r 177 T present include Sc, v Au, Ta, Eu, La. Lu,

153 175 Sm, and Yb . Thus, of the elements of interest, only

Au could be determined nondestructively , even in principle;

but, because the actual Au concentration level was in some

cases only 5-10$ of the nominal (5 ppm instead of 50; 20

; ppm instead of 500), separation of the Au was necessary to

I obtain the degree of precision required for certification.

24 Because of the extremely high Na levels present immediately after irradiation, and the resulting high radiation

; fields (>100 r/hr per sample at 1 cm) encountered in transferring the irradiated samples to storage containers, it was

! desirable to minimize the number of irradiations by determining

all the elements of interest in the same sample. The following

procedure was therefore developed and utilized.

The irradiated glasses were dissolved in HF-HCIOk and

I carried to fumes of HCIOk. The residue //as taken up in 6N

HC1 with a few drops of HNCU and Au was removed by extraction

with isopropyl ether. The ether phase was adjusted to a standard volume and gold was determined by counting with a

' 47-cc Ge(Li) detector and a multichannel analyzer.

The aqueous phase was evaporated nearly to dryness and

the residue taken up in . 2N HCIOk. Silver was separated by

extraction into CHC1„ of the dibenzyldithiocarbamate . Any

entrained Sc and rare earth activities were removed by

washing the CHCl^ phase with a small amount of . 2N HCIO^.

The CHC1- phase was adjusted to a standard volume and silver

was determined by counting with a 3 in x 3 in Nal(Tl) detector and a multichannel analyzer. This method gave a silver

fraction of high radiochemical purity, in contrast to other

methods tried, including precipitation of AgCl and of Ag

metal which give products contaminated with Sc, Ta, or various

rare earths . The remaining aqueous phase, ^0 . 2-0 . 5N HCIO^ and con46

taining the cobalt, had too much activity from Sc and

57
152 various rare earths, including Eu to be counted directly.

Therefore the bulk of the Eu and heavy rare earths were

removed by extraction with HDEHP . Sc was removed by a second

HDEHP extraction after acidification with HNCU. Cobalt was

then determined in the aqueous phase by counting with a 47-cc

Ge(Li) detector to separate the Co peaks from those of

Ta and remaining traces of Eu.

In all cases concentrations were determined by comparison

with standards (Co, Ag, and Au metal foils) which had been

irradiated simultaneously with the samples, dissolved, and

made up to the same volume for counting.

The results obtained are shown in Tables 25, 26, and

27. The scatter in the Ag and Co results for the 500 ppm

glass may be due in part to the presence of a small amount

of precipitate, probably TapCv, which could be observed when

the samples were taken up in . 2N HClCv following the

evaporation to dryness after Au extraction. This precipitate

was not present in the more dilute 50 ppm and 1 ppm glasses.

Table 25. Results of analysis of 1 ppm glass (SRM 615).

Sample Co, ppm Ag, ppm Au, ppm

Rod # 2A 0.724 0.470 0.591

0.565

Co, ppm

.724

.725

.738

.741

.723

.729

.709

.744

.730

711

726

729

0. 727

Oil

Ag , ppm

.470

.455

.469

.468

36A 0.738 0.469 0.591

0.572

40A 0.723 0.460 0.557

0.590

69A 0.709 0.460 0.547

0.446 0.560

76A 0.730 0.596

0.591

114A 0.726 0.541

0.559

x ± S 0.727 0.461 0.572

±0.008 ±0.019
Table 26. Results of analysis of 50 ppm glass (SRM 613).

Sample

Rod # 2

Co,

37.1

ppm

36.3

35.2

Ag, ppm

25.0

24.0

Au , ppm

5.62

5.57

37 36.8

35.9

34.9

34.1

21.8 5.44

5.41

5. 4l

44 35.6

36.1

35.4

33-7

23.3 5.62

5.38

5.26

68 37.6

35.6

36.5

35.0

24.1

25.2

5.31

5.30

5.20

75 35.1 34.8

35.8

23.0

22.7

5.36

5.14

114 34.1 35.4

35.8

21.4

23.8

5.43

5.14

x ± S 35.5+1 .0 23.4±1.2 5.38±0.l4

Table 27. Results of analysis of 500 ppm glass (SRM 611

Sample Co 3 ppm

Rod § 2 367

13 408

18 393

48 384

56 370

66 368

78 360

106 372

Ag 3 ppm

221

255

226

233

225

361

Au > PPm

22.9

25.3

24.0

24.9

23.6

23.8

x ± S 378±16 237±17 24.1±0.9

(B. A. Thompson)

59
3 . Survey of SRM 20 ppb Trace Elements In Glass

A survey analysis was conducted on the trace elements in

glass to which 20 ppb of 6l trace elements had been added.

The procedure used was that previously reported for the analysis of base glass; i.e.

1) irradiation in RT-3 for 15 minutes

2) washed with 1:1 HNO

3) dissolution with HClGv and HF

4) removal of Na-24 with HAP column

5) 30 and 720 minute count on a 47 cc Ge(Li) detector

coupled to a 2048 channel analyzer

6) peak location and area calculation via the ALSPIS

computer program written by H. P. Yule.

Gold, europium, copper, lanthanum, antimony, gallium,

and scandium were observed in all samples . The results are

given in table 28. Scandium results have a greater than

Table 28. Glass impurities.

Sample Au Er Cu La Sb Ga Sc

2A 0.169 0.015 0.827 0.032 0.081 0.233 0.026

2A 0.205 0.013 0.751 0.028

33 0.176 0.016 0.781 0.033 0.077

33 0.197 0.013 0.715 0.038

43 0.181 0.015 0.808 0.039 0.080

43 0.178 0.016 0.922 0.035

43A 0.206 0.016 0.831 0.035 0.080

43A 0.138 0.014 0.797 0.028

51 0.180 0.017 0.858 0.038 0.076

51 0.175 0.014 0.827 0.038

51A 0.165 0.015 0.758 0.035 0.071

51A 0.184 0.016 0.831 0.036

76 0.188 0.016 O.786 0.035 0.080

76 0.147 0.013 0.799 0.040

76A 0.188 0.017 0.827 0.034 0.081

76A 0.189 0.015 0.035

Recommended

Value

0.179 0.015 0.808

S ±0.018 ±0.0013 ±0.049

ts//n±0.010 ±0.0007 ±0.027

0.253 .020

0.205

0.230 .022

0.182

0.258 .023

0.257

0.242 .023

0.186

0.238 .028

0.245 .033

0.212

0.243 .032

0.243

0.035 0.078

±0.003 ±0.003

±0.0018 ±0.003

0.231

±0.025

±0.014

0.026

0.005

0.004

60
average uncertainty due to the low solubility of scandium

chloride in 8-12N HC1. Gold values have been corrected for

self absorption in the standard.

(H. L. Rook and W. D. Kinard)

4 . Survey of SRM One ppm Trace Elements in Glass

A survey analysis was conducted on the trace elements in

glass to which 1 ppm of 6l trace elements had been added.

The procedure was the same as for that of the 20 ppb glass

reported above.

Gold, europium, lanthanum, antimony, gallium, scandium,

and copper were observed in all samples. The results are

given in table 29. The gallium results are only significant

to ^±10% since europium in the 1 ppm glass increased by a factor of 10 relative to gallium when compared to the 20 ppb

glasses. This necessitated the use of a less abundant photopeak for analysis

.

Table 29. Analytical Results for 1 ppm Glass.

Concentration (ppm)

Sample

No. Au

0.436

La

0.842

Eu

0.998

Sb

1.06

Ga

1.4

Sc

0.565

Cu

2A 1.56

36A 0.438 0.842 1.001 0.98 1.5 0.579 1.61

40A 0.408 0.826 0.979 1.09 1.2 0.603 1.44

69A 0.332 0.835 0.979 1.13 1.3 0.597 1.65

76A 0.308 0.810 0.972 1.12 1.3 0.592 1.69

114A 0.383 0.842 1.015 1.04 1.2 0.596 1.64

76 0.369 0.839 1.009 1.05 1.2 O.606 1.52

Ave . 0.382 0.834 0.991 1.07 1.3 0.591 1.59

Ts//n 0.046 0.011 0.016 0.048 0.11 0.013 0.080

(H. L. Rook)

61
5. The Determination of Manganese in Trace Elements in

Glass Samples

Manganese was determined in glass disc samples at NBS

as part of the SRM certification. Manganese is one of 61

different elements which will be certified in concentrations

ranging from 1 to 500 ppm.

a

Experimental

The glass discs were weighed and encapsulated in polyethylene tubing. The weights ranged from 0.40 gram to 0.80

gram. Copper foils were attached to each sample and standard

for flux normalization. Manganese metal (99-99+ purity)

was the standard used. The samples and standard were irradiated at appropriate times in the NBSR pneumatic tube facility

11 -2 -1

at a flux of ^5x10 n.cm sec

The discs were washed with 1:1 nitric acid for 1 minute

and dried with ethyl alcohol. The discs. were then dissolved

with HF-HCIOk 1:1 in a Teflon beaker. The solution was

passed through a hydrated antimony pentoxide (HAP) column to

remove the sodium-24 activity [18, 8]. An additional 5 ml of

concentrated HC1 was used to wash the column. The eluent

was made 7M in HC1 and extracted with isopropyl ether to re72

move Ga which has a 0.840 MeV peak which interferes with

the 0.845 MeV peak of manganese-56 . The aqueous phase was

made to a constant volume and counted on a 3 in. x 3 in.

Nal detector with 400 channel analyzer. The standards were

dissolved and counted at the same geometry as the samples.

In the case of the 500 ppm glass, the additional activities

made determination of the Mn difficult without further

separations. In this case the sample was dissolved in hot

1:1 HC1CK-HF, taken to incipient dryness and manganese precipitated as the dioxide from 6M HNO q with KBrO~. The

manganese concentrations were calculated using the standard

comparator method.

bResults

The results are given in table 30.

62
Table 30. Manganese concentrations in trace elements

in glass samples.

Sample Disc Amount Found Sample Disc Amount Found

No. No. (ppm) No. No . (ppm)

Rod

Base 61A .512 SRM 611 56 469

glass

Rod

61A .512

.519

31A .502

.499

117A .461

.461

107A .462

.458

41A .413

.417

8A .492

.468

87A .518

• 513

.522

131A .463

.478

.465

56 477

48 466

48 487

2 479

18 469

66 484

106 477

Average 476±.0006 a

Average 470±.0l4 a ppb

Uncertainties are ts//n at 95% C. L.

(T. E. Gills)

6 . Intermediate Purity Sulfur

A survey analysis was conducted on the intermediate purity

sulfur with emphasis on sample homogeneity. Replicate 0.5

gram samples were taken from each batch by placing a sulfur

lump in a new polyethylene bag and fracturing it with a soft

blow from a plastic beaker. Enough pieces to make 0.5 gram

were sealed in precleaned medical grade polyethylene tubing,

irradiated in RT-3 of the NBSR for 15 minutes. The samples

were transferred to clean polyvials and counted on the 47cc

63
Table 31. Impurities found In intermediate

purity sulfur (ppm by weight).

LOT - 1

Sample Mn As Laa Cu Na

1 can 1

-4

0.7x10

-4

2.3x10

0.033 0.026 N.D. 0.10

1 can 1 0.023 0.015 N.D. 0.26

1 can 2 2.4x10"^ 0.025 N.D. N.D. 0.03

2 can 2 1.5X10" 4

0.030 N.D. N.D. 0.10

1 can 3 6.5xl0

_l1

0.050 N.D. N.D. 0.21

2 can 3 6.1xl0

_i|

0.042 N.D. N.D. 0.36

1 can 4 l.OxlO" 3 0.01 0.27 0.035 0.10

2 can 4 1.5xlO~ 3 0.19 0.14 0.037 0.10

1 can 5

2 can 5

-4

1.5x10

-4 1.3x10

0.029

0.027

0.19

0.15

0.031

0.55

0.08

0.03

p

b Limit of detection under contitions used ^0.005 ppm. Limit of detection under conditions used ^0.001 ppm

N.D. = not detected.

Ge(Li) detector coupled with a 2048 channel analyzer. The

results are given in table 31.

Impurities found to be present were manganese, arsenic,

lanthanum, copper, and sodium. Cans 3 and 4 were found to

have approximately a factor of 3 and 10 higher concentrations

of trace elements respectively. The large variation in

results indicated a high degree of inhomogeneity . Because of

this, a second analysis was carried out where samples were

segregated between pieces taken from the smooth surface,

which cooled rapidly on a plastic sheet and pieces taken from

the interior of the sample. The results are shown in table

32. These results clearly indicate a segregation of impuri64
Table 32. Variation in impurities found - surface

vs. interior of sample (ppm by weight).

Sample Mn

1 0.7xl0 -3

2 0.9xl0~ 3

3 3.1xl0" 3

4 0.9xl0~ 3

La Cu As

Smooth Su rface

0.9xl0~ 2 4xl0~ 3 0.014

N.D. N.D. 0.015

l.lxlO

-2

3xl0 -3 0.008

0.7xl0~ 2

N.D. 0.017

0.8xl0" 2 1.8xl0~ 3 0.013

Int<5rior

8xl0~ 2 l.lxlO" 2 0.008

5xl0

-2

3.2xl0~ 2 0.019

8xl0~ 2 1.2xl0

-2

0.005

2xl0" 2N.D. 0.010

5.7xl0

-2

1.5xl0

-2

0.011

Average 1.4x10

1 5.6xl0~ 3

2 3-3xl0~ 3

3 7.0xl0" 3

4 1.5xl0~ 3

Average 4.4x10

N.D. = not detected.

I ties, except arsenic, to the interior of the sulfur cake.

j This may have been due to an internal "zone refining" effect

as the sulfur cooled.

(H. L. Rook)

7 . Potassium Chloride Single Crystal

A survey analysis was conducted of a high purity KC1

crystal to determine if impurities from its plastic container

were contaminating the crystal surface. Two different procedures were used to detect surface contamination. First, a KC1 crystal was transferred and sealed into medical grade

polyethylene which had been washed in 1:1 nitric acid, rinsed

| with deionized water, followed by absolute ethanol and drying

The sample trnasfer from package to polyethylene tubing was

65
conducted without contact with any foreign implements. The

sample was irradiated in KT-H for 5 minutes. Following

irradiation, the sample was etched in 3 successive portions

of deionized water, each removing about 50 mg of KC1. The

solutions were counted in a standard geometry on the 47cc

Ge(Li) detector for 30 minutes after a 4 hour delay to allow

the decay of chlorine-38. All solutions contained only activity derived from potassium and bromine. The weight percent

bromine was the same in all three fractions thus it was concluded to be a bulk impurity rather than a surface contamination . The second and more sensitive procedure was to open a package with a scalpel along one edge, remove the KC1 crystal

and reseal the package with polyester tape. The package was

irradiated in RT-3 for 25 minutes, the tape removed, the KC1

crystal reinserted, vigorously rubbed with the inside of

the package, removed and counted. The only contamination

found other than potassium and chlorine, was sodium in the

level of ^10 ug. It was felt from these results that no

significant contamination of the KC1 crystal was occurring

from contact with the packaging material.

(H. L. Rook)

8 , Determination of Indium and Copper in SRM High

Purity Gold Wire by Neutron Activation Analysis

Samples of high purity gold wire were received from the

Office of Standard Reference Materials for analysis of impurities. Information available about the samples, from

resistivity measurements and survey analyses, indicated

that the two trace elements indium and copper were present

in the gold at sub-ppm levels. If possible, activation

analysis was to be used to obtain a quantitative measurement of these impurity levels for certification.

The gold matrix presented a formidable problem in this

analysis. Gold, with a 98.8 barn cross-section for thermal

66
neutrons and a 64.8 h half-life produced approximately 20

198 millicuries of Au for a 50 mg sample irradiated for 30 m

12 -2 —1

at a thermal neutron flux of 8x10 n.cm sec . In contrast,

_7

a copper impurity of 0.1 ppm would produce only 5x10 as much radioactivity as the gold. In addition, both the

indium-ll6m (T , =54m) and the copper-64 (T /p =12.9 h) have

considerably shorter half-lives than the gold. As a result

of these unfavorable analysis conditions, a radiochemical

separation of the copper and indium was required. This

separation not only had to be rapid, in order to count the

short-lived Indium radioisotope, but had to be very clean,

h

6

5 with a decontamination factor of at least 10 and preferably

>10

A radiochemical separation which would fill these requirements was developed. It consisted of spontaneous

deposition of the gold matrix on a metallic silver-asbestos

column, with a final purification by precipitation of the

copper and indium cyanoferrates

.

a. Experimental

The samples as received were in the form of 1.4 mm diameter wire. They were separated into short sections and

flattened between two clean lucite blocks to reduce neutron

self-shielding effects. Individual samples weighed approximately 50 mg and were irradiated along with suitable standards and flux monitors for 30 minutes in the NRL in-core

12 -2 -1 facility at a thermal neutron flux of 8x10 n.cm sec

After irradiation, the samples and standards were returned

to NBS , etched in aqua regia to remove exterior contamination,

and then dissolved for the analysis with addition of copper

and indium carriers. After dissolution, perchloric acid was

added and the solutions taken to fumes of perchloric acid to

remove the chloride ions. The silver column was prepared by

adding a metallic silver slurry to purified asbestos and

thoroughly mixing in a blender. This suspension of silver

metal on asbestos was then poured into a filter chimney and

67
kept under water. After preparation of the column, the concentrated perchloric acid sample solution was diluted and

passed twice through the silver-asbestos column. The column

was then washed once with dilute perchloric acid. This procedure provided both a 10 decontamination of the filtrate

from gold-198 and a quantitative recovery of the copper and

indium in the eluate.

In a final separation step, to remove the last traces of

gold and to remove a small amount of sodium-24 contamination,

the copper and indium were precipitated as their cyanoferrates

To do this, the eluate was pH adjusted and a saturated solution of K|,Fe(CN)^ added. The highly insoluble copper and

indium cyanoferrates were centrifuged, the supernate decanted,

and the precipitates then washed with dilute K^PeCCN) r solution.

The copper and indium precipitates were counted while

still in the centrifuge tubes, on a 3x3 Nal(Tl) scintillation

detector with a 400 channel PHA . Half-lives of both copper

and indium were measured to verify isotopic purity, and appropriate calculations and corrections applied to the data.

b . Results and Discussion

The results of these analyses are found in table 33.

These results include a calculated neutron self-shielding

correction of almost 20%

The analytical results have rather large uncertainties

(±22% for copper and ±26% for indium) due primarily to the

extremely unfavorable matrix involved. In spite of the

difficulties encountered, the agreement between these values

and other analytical techniques was satisfactory and the

high-purity gold was certified for its copper and indium

impurity concentrations [20].

68
Table 33- Analyses for copper and indium in SRM

gold wire.

Element Determined (ppm)

Sample Copper Indium

1 0.12, 0.11 0.005

0.012

2 0.25 0.014

0.014

0.010

0.012

0.006

0.010

0.0027

hNot determined.

95% confidence level.

3 N.D. a

4 0.11, 0.21, 0.15

0.18, 0.16

Mean 0.16

ts//n b ±0.034

(W. D. Kinard, D. A. Becker, and P. D. LaPleur)

9 . The Determination of Cobalt in Cobalt-Aluminum

Flux Wires

The NBS Standard Reference Material program is certifying

cobalt-aluminum wires for use in neutron flux determinations

in the nuclear industry. The wires are approximately 10 mil

diameter and contain 0.1% cobalt. These wires were analyzed

by the Activation Analysis Section for their homogeneity and

cobalt content.

a. Experimental

The wires were cut into 1.5 inch lengths and weighed.

The average weight of each sample was approximately 14 mg.

The wires were then encapsulated in clean quartz. Attached

to each sample was a 5.0 mg copper foil to monitor the flux

variation. Cobalt metal (99-9+%) was also encapsulated in

69
Table 34 „ Analysis results for cobalt In cobalt

aluminum wire.

Cobalt Average % Cobalt Average %

Samp le No. #_ Cobalt Sample No. % Cobalt

.118

.117

.115

.117 + .004

.117

.118

.116

.117 + .002

.118

.122

.119

.120± .05

.117

.117

.116

.117± .001

.117

.118

.118

.117 + .002

.116

.117

.118

.117+- .002

.119

.121

.116

.119± .006

.117

.118

.117± .006

10

11

12

13

14

15

.117

.118

.118

.118± .001

.116

.117

.119

.117± .004

.118

.114

.116

.116± .005

.119

.118

.119

.119± .001

.116

• 119

.115

.117± .005

.118

.115

.116

.116± .004

.110

.114

.109

.111± .006

Errors expressed are ts/n at the 95% confidence level.

quartz and was irradiated along with the samples. The

samples and standards were irradiated for approximately five

hours at a neutron flux of 8x10 n.cm~

2

sec~

1 in the glory

tube of the Naval Research Reactor in Washington, D. C.

After irradiation the samples were allowed to stand for

three weeks so that the sodium-24 activity produced by the

27 24 Al(n,a) Na reaction on the matrix could decay. Sodium-24

70
has a fifteen hour half-life and a I.36 MeV gamma ray that is

very close to the 1.33 MeV of Co . After the sodium-24

activity had decayed, the samples and standards were counted

using a 3 in. x 3 in. Nal(Tl) scintillation detector and a 400 channel pulse height analyzer. The cobalt content of

the samples were calculated using the standard-comparator

method. The results were normalized for flux variations

using the flux monitors. The results are given in table 34.

b . Results and Discussion

The results showed no significant difference between

samples 1-14. Sample 15 is significantly lower in cobalt

content. This result was confirmed by spectrophotometry.

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

10 . Analysis of Cast Iron Steels for Tellurium

SRM No. 1140 (Cast Iron Steel) was analyzed for its

tellurium content using destructive neutron activation analysis [21, 22]. The procedure utilizes the nuclear reaction:

Te 130

(n, Y )Te 131 *' ^

2 5m

a. Experimental

Individual cast iron samples of ^145 mg were weighed and

encapsulated in polyethylene tubing. To each sample was

attached a copper flux monitor to measure flux variation.

IS The samples were irradiated for ^1 hr at a flux of ^5x10

-2 -1

n.cm sec . After allowing to decay for 12 hours, the iron

samples were dissolved in a closed system and the radioactive

iodine distilled and precipitated as Agl . Iodine-125 was

used as a tracer for yield measurements. Standards were

small (^1 mg) particles of 5-9 's tellurium metal.

b

.

Results

The results are shown in table 35- The cause of the

lack of precision is unknown at present, but may be due

71
Table 35. Results of tellurium analysis in SRM cast

iron steel.

SRM No. Te Concentrations % Average

1143 .022

.019

.020

1143a .016

.017

.016

1144 .022

.017

.019

1144a .017

.026

.022

1147 .011

.013

.020

.015

1147a .029

.036

.032

1148 .018

.013

.015

1149 .011

.011

.011

either to inhomogeneity in the samples or to factors not

defined in the analytical technique.

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

11. Activation Analysis of SRM Steels Using Group Radiochemical Separations

Concentrations of several minor constituents in SRM steels

1263, 1264 and 1265 have been determined by activation analysis using the NBS Reactor and group radiochemical separations reported previously [23]. For SRMs 1263 and 1264

the analyses were for acceptance testing of the sample submitted by the manufacturer, and for 1265 the purpose was

certification. The results are shown in table 36.

72
Table 36 Results of analysis c (% by weight unless c

>f SRMs 1263, 1264, 1265

therwise noted).

Element

Cu

As

1263

0.0093

1264 1265(365)

0.00593 ±0.00029 a 0.000224±0 .000010

Co

Cr

W

Sb

As

La

0.0279

0.00153

0.00052

0.00103

0.0299

0.0000965

0.000795

0.00742 ±0.00038

O.OO698 ±0.00063

0.000033±0. 000007

<0 . 5 ppm

<0.02 ppm

<0.1 ppm

Ta

Al 0.253

<0.5 ppm

Limits quoted are ts//n for the 95% C.L.; n=6.

SRM ductile irons 1140, Il4l, 1142, 1140A, 1141A, and

1142A were analyzed for Cu, As, and Mo by neutron activation

Cu and As were determined by counting with a 47cc Ge(Li)

detector after extracting Fe and Sb as the chlorides with

isopropyl ether. Mo was determined by a double extraction

99m with HDEHP , using Tc as a measure of the Mo present

[24]. The results obtained are shown in table 37.

Table 37

SRM No.

1140

1141

1142

1140A

1141A

1142A

Results of analysis of SRM ductile irons

( % by weight )

.

Cu As Mo

0.1084±0.00l8 a 0.0718 ±0.0022 0.1003±0.0021

0.2473±0.0065 0.0360 ±0.009 0.0478±0.0028

1.169 ±0.016 0.00503±0. 00010 0.0213±0.0006

0.1013±0.0019 0.0782 ±0.0008 0.l620±0.0048b

0.2331±0.0082 0.0213 ±0.0008 0.0439±0.0053b

1.187 ±0.011 0.00456±0. 00022 . 0235 + . 0012 b

^Limits quoted are ts//n for the 95% C. L. ; n=6

n=3.

73
Molybdenum was determined in SRM lOlf using the HDEHP

double extraction method. The results obtained on two different samples of lOlf were:

101f-l 0.00703±0. 00031 % Mo

101f-4 0.00668±0. 00021 % Mo

where the limits are ts//n for the 95% C. L.; n=6.

Arsenic was determined in SRM 122e (cast iron car wheel)

by neutron activation analysis. Any potential antimony interference was removed by extraction with isopropyl ether from

^7N HC1 in the presence of NaBrO . The value found was

0.01766±0. 00010 % As

where the limits are ts//n for the 95% C. L . ; n=6

(B. A. Thompson)

12. The Determination of Cobalt in SRM Steels by

Neutron Activation Analysis

Cobalt was determined in NBS steels lOlf-. and lOlfj. by

nondestructive neutron activation analysis. The steels are

part of the certification program at the National Bureau

of Standards

a. Experimental : The steel samples were weighed

(^100 mg ea) and encapsulated in polyethylene tubing.

Copper foils were attached to both samples and standards

to monitor flux variation. The samples were irraditaed

13 -2 -1 for 10 minutes at a neutron flux of 1.3x10 n.cm sec

The samples were allowed to decay for 48 hours before

counting. The counting was done on a 47cc Ge(Li) detector

in conjunction with a 4096 channel analyzer. The flux

monitors were used to normalize results.

b

Results

The results of the analysis are given in table 38.

NBS certified SRM Steel No. 1166 was analyzed concurrently

to verify the accuracy of analysis.

74
Table 38. Results of analysis for cobalt

(% by weight )

.

Sample

101f

1

101f

4

Amount found Average

.0873

.0851

.0869 .0868±.001

.0877

.0871

.0848

.0875

.0840

.0877 .0863±.001

.0853

.0883

(T. E. Gills)

13 . Determination of Copper and Cobalt in SRM Steel

1265 B-l

SRM Steel No. 1265 B-l was analyzed for copper and cobalt

using destructive neutron activation analysis. The sample

contained as many as 13 elements which when irradiated in a thermal neutron flux produce a gamma spectrum too complex for

resolution by a Ge(Li) detector. Therefore, a separation

of copper and cobalt from the matrix was necessary.

a. Procedure

The steel chip samples were weighed, (^100 mg ea.) encapsulated in polyethylene tubing and irradiated for 10 minutes

.

To each sample 10 mg copper foils were attached as flux

monitors. The samples were allowed to decay for 24 hours to

56 reduce Mn activity. The samples were then dissolved in

HC1CK-HN0 and Cu and Co were separated using a solvent

extraction procedure [25]. The separated fractions were

counted on a 47cc Ge(Li) detector in conjunction with a 4096

channel analyzer. Concentrations were determined using the

standard comparator method. Flux monitor foils were used to

normalize results.

b

.

Results

The results are given in table 39.

75
Table 39- Results of activation analysis of SRM 1265 B-l.

Average ± ts//n

Element % Found (95% C .L.

Copper .00597

.00605

.00632

.00628 .006ll±. 00014

.00603

.00602

.00584

.00637

.00614

Cobalt .00724

.00715

.00771

.00740 .00747± .00020

.00758

.00758

.00735

.00749

(T. E. Gills)

14 . Nondestructive Determination of Manganese and Vanad- ium in Iron and Steel Standard Reference Materials

Nine iron-based SRM's were analyzed for their manganese

and vanadium content. The method used was nondestructive

neutron activation analysis. Besides the two elements of

interest, the SRM had additional added elements to a maximum

of 13 per SRM, averaging about 10 certified elements. The

elements range in concentration from <_.!% to" over 3% (except

for major constituents such as chromium which is ^20%).

Similar samples have been priviously analyzed for manganese and vanadium in this laboratory [26, 27]. In most

cases the procedures were essentially the same. Therefore

only the general procedure will be discussed here.

a. Experimental

Samples were analyzed as received. Approximately 100

mgs each of the steels were sealed in polyethylene snap-cap

vials . The standards were prepared by pipetting solutions

of manganese and vanadium onto filter paper inside the vials

76
and allowing-t dry. The vials were then heat-sealed. Copper

foils were attached to both sample and standard to normalize

the results

.

The procedure used was as follows:

(1) The sample or standard was irradiated for 3 sec in

pneumatic tube RT-4 of the NBS reactor.

(2) Allowed the sample to decay for 3 minutes. During

this time the flux monitor was removed and the

sample positioned on the detector.

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

4.0 minutes live time with the clock time recorded

and used for decay correction.

The peak areas of both manganese and vanadium was integrated

using Covell's method [28]. The concentrations of each sample

was then calculated using the standard-comparator method. Flux

variations from sample to sample were corrected using the flux

monitors. Shelf-shielding and gamma-ray attenuation corrections

were calculated and found to be negligible for these conditions.

b . Results and Discussion

The analysis of the eleven SRMs can be found in table 40.

Table 40. Results of manganese and vanadium analyses

of iron and steel Standard Reference Materials.

SRM No. Of Average % No. of Average %

Sample Samples Mn Found Samples V Found

lOlfl 6 .0904±.0013 a

6 .0284 ±.0003

101f4 6 .0905±.0026 6 .0287 ±.0007

342a _ — 5 .0192 ±.0009

1261 12 .6845+. 0101 6 .011321. 00021

361 12 .6803±.0054 6 .011321 .00028

1265 B-l 43 .0066+. 0002 9 .00058±. 00002

1265 8 .0074±.0002 9 .00054±. 00002

1138 4 .4393+. 0140 4 .0225 ±.0020

1139 4 .9662±.0l84 4 .2405 1.0081

362 6 1.024 ±.012 6 .0394 ±.0004

1262 6 1.030 ±.009 6 .0392 ±.0007

Uncertainties are ts//n at 95% C. L.

(T. E. Gills)

77
15 . The Multielement Analysis of SRM Steels 1261 and

and 36l by Destructive Neutron Activation Analysis

As a part of the SRM program at NBS, steel SRM's 1261

and 361 were analyzed for final certification. These steels

contain approximately 13 elements of which 8 were analyzed

by neutron activation analysis (W, Cr, Cu, Cg, As, Sb , La

and Mo)

a. Experimental

Six samples of each of the steels were irradiated for one

hour in RT-3 of the NBS Reactor. The samples were allowed

56 to decay over night to reduce the Mn activity. The samples

were then dissolved and separated into three groups using

a solvent extraction procedure [25]. W and Mo were determined by counting the a-benzoinoxime CHC1„ fraction. T_e

other elements were determined by counting the ether extraction and aqueous fraction. The counting was done on a 47

Li(Ge) detector in conjunction with a4096 pulse height

analyzer.

b

Results

The results obtained are shown in table 4l.

Table 4l. Results

steels

of multielement

1261 and 361.

analy sis of SRM

Element /Samp

e 1261 361

Cobalt .0325 ±.0007 a .0329 ±.00l6 a Copper .0432 ±.0016 .0433 +.0004

Chromium .624 ±.060b

.637 ±.017b

Arsenic .0128 ±.0009 .0145 +.0008

Antimony .00270 ±.0002 .00259±.0002

Lanthanum .000378±.000030 .00109+. 00008

Molybdenum .1531 ±.0180 .1582 ±.0045

Tungsten .0107 +.0021 .0107 ±.0009

Limits quoted are ts/Vn at the 95% confidence level

n=6 except where noted.

b n=4 (T. E. Gills)

78
E . Special Analyses

1 • Determination of Copper Homogeneity In a Copper-Lead

Alloy

In an attempt by the Activation Analysis Section to obtain an ideal flux monitor (see also Section 3_c-4s this

report) a bid was submitted by a private company to fabricate

a homogeneous alloy of 0.01% copper in lead. The specifications stated that the absolute copper concentration was not

critical, but that the homogeneity must be at least ±1/2%.

The lead diluant was to be at least 99-999% pure, and samples

of the lead to be used were to be submitted to us for analysis

before alloying with copper.

The successful production of a homogeneous copper-lead

alloy on a small scale had already been made in this laboratory [29]. In this previous work 0.1% copper foils with

an S, of 1.08% (n=26) were produced. It was expected that

with professional facilities a more homogeneous alloy could

be produced.

a. Experimental

The foil was obtained as a series of long narrow sheets

of ^5 mil thickness. After cleaning the foil surface with

ethyl alcohol, samples for analysis were taken from various

portions of the sheets . These samples were irradiated along

with a copper standard for 15 minutes in the RT-3 pneumatic

facility of the NBSR. Flux normalization was made for both

samples and standards.

64 After verifying the radiochemical purity of the Cu

activity using a Ge(Li) detector and 4096 channel PHA, the

samples and standards were counted on a 3 x 3 Nal(Tl) detector

and 400 channel PHA.

b. Results and Discussion

The results of these analyses are much too numerous to

list. However, at no time did the foils received meet specifications. In a number of cases, the homogeneity from one

79
area would be quite good 00.5% S ); however, over the whole

batch the foil was quite inhomogeneous (up to 3.0% S , ) . Reruns of our in-house produced foil showed significantly better

homogeneity than the commercially produced foil.

In addition to the problem of the copper inhomogeneity

reworking of the alloy by the company introduced a relatively

large, Inhomogeneous manganese impurity into the foil. This

impurity ranged from ^8 to 24 ppm, and immediately after

irradiation produced a greater radioactivity due to Mn

64 than that of the Cu . Thus, the alloy was determined to

be unsatisfactory for use as a flux monitor material.

(W. D. Kinard and D. A. Becker)

2 . Analysis of Neodymium Laser Glass by Neutron

Activation

The Activation Analysis Section was asked by the Inorganic Materials Division to carry out analyses on five

different neodymium laser glasses as part of an overall

program of characterization of these glasses. The glasses

are very complex mixtures with, In some cases, as many as ten

principal constituents (Si, Na, Ba, K, Al, Nd, Sb , Zn, Li, Ti),

For activation analysis the constituents representing the most

serious interferences to the measurement of other activities

are sodium, antimony (1% or more in 4 of the glasses), and

neodymium (3-5% in all 5 glasses). Neodymium is especially

aggravating since on irradiation it yields three radionuclides:

151Nd, with a 12 minute half-life; 9 Nd, with a 1.8 hour

half-life; and 7 Nd with a 11 day half-life. 151Nd and

149 Nd also decay to form radioisotopes of p 'omethium with

half-lives of 28 hours and 53 hours. All th se nuclides have

very complex decay schemes and emit many gamm -rays.

To try to obtain the maximum amount of inl rmation with

a reasonable expenditure of effort, three sets of irradiations

were carried out. For each of the glasses a small sample

(10-50mg) was irradiated for 2 seconds and used to determine

80
Na, K, and Sb . A larger sample (100-200 mg) of each glass

was Irradiated for 1 minute and dissolved in HF-HCIO^. After

fuming to eliminate HF, the residue was taken up in 9N HC1

and most of the antimony was removed by extraction into isopropyl ether. Sodium was then removed by passing the aqueous

phase through column of hydrated antimony pentoxide. The

effluent was examined for low-level constituents of intermediate half-life using a 47-cc Ge(Li) detector. Those observed included La, Pr, Eu, and Sm. Common impurities such

as Cu and Ga were not observed.

A third set of samples (100-200 mg) was irradiated for

15 minutes and set aside for two weeks to allow all shorterlives activities to decay away. After dissolution and

extraction of antimony as before, the aqueous phase was

counted and Nd, Ce, and Ba were determined. (Glass "D" did

not require these chemical separations since it had very

low concentrations of Na and Sb .

All the samples were irradiated in pneumatic tube RT-3 of

IS -2 -1 the NBSR at a flux of about 5x10 n.cm sec. . The maximum

irradiation time-sample size combination was fixed by the

24 122 124 147 Na, Sb- Sb , and Nd activities produced. All the

samples were specially cut 1-mm thick slices. With samples

this thin, neutron self-shielding was not a problem even for

the "D" glass with lk% LipO. For this latter glass irradiation time was not limited by Na or Sb because these were

quite low, but by Nd and by heating of the sample during

irradiation by absorption of tritons and alpha particles

from the reaction Li(n,a) H. All the samples were rinsed in

acetone and in dilute HN0_ after irradiation to remove surface

contamination. Concentrations were determined by comparison

to standards irradiated under the same conditions.

The results obtained are shown in Table 42. For most

elements the standard deviation is estimated to be about ±5% - For the lower reported values of Sm there is severe interference from Nd and these values are probably no better than

81
Table 42. Results of activation analysis of neodymium

glasses

H A1t llgH HQtl "D'

Na

2 0, % 7.83 3-19 6.39 0.044 6.85

K 0, % 13.8 16.6 17.1 <0.04 9.21

SbpO , %

Nd

2 3

, %

BaO, %

0.86 2.5^ 0.93 0.0015 0.90

5.84 3.53 5.73 3.52 5.52

3.27 3.46 2.83 <0.3 5.06

Ce0

2

, % 0.005 <0.006 <0.006 0.52 <0.006

La

?

, ppm

Pr

?

, ppm

10.2 17.8 7.8 <0.8 <0.9

<3 <3 8.4 <2 <3

Eu

2

, ppm

SrruO,., ppm

0.6

0.2

<0.1

0.2

1.1

2.2

<0.5

<0.03

29.7

0.37

±20%. For the lower reported values of Eu and Ce the signalto-background ratio was poor and these values also are probably

no better than ±20$.

(B. A. Thompson)

3 . Determination of Trace Elements in Ruby Laser

Crystals

At the request of the Inorganic Materials Division a number of ruby laser crystals were analyzed quantitatively

for trace elements by neutron activation. The analysis is

quite difficult for other methods such as emission spectroscopy and spark source mass spectrometry because the ruby

(a-AlpO_ with <_0.1% Cr ? CO is such a good insulator and thus

does not spark. A1

? 0^ is almost an ideal matrix for activation analysis since the longest lived radionuclide produced

24 from the matrix is 15-hour Na formed by the reaction

?7 ?4

'Al(n,a) Na.

82
a-Al ? 0_ is extremely refractory and so all the analyses

were carried out nondestructively . The crystals were cut

to a standard geometry (see fig. 22) before irradiation

and were irradiated in pneumatic tube RT-4 of the NBSR which

has a high ratio of thermal-to-fast neutrons. This facility

Figure 22. Ruby crystal as cut for analysis

13 -2 -1 has a flux of ^10 n.cm sec. . After irradiation and

p o decay of " Al 3 the samples were counted at several times

with a 47-cc Ge(Li) detector. Concentration levels for the

observed trace elements were determined by comparison to

standards irradiated under the same conditions and dissolved

in a few drops of appropriate solvent to give the same geometry as the samples. Three sets of crystals were analyzed

and the results are shown in tables 43, 44, and 45. The

results listed in table 45 were obtained on current production material. Detection limits for all the elements which

could have been detected if present under these conditions

were determined so that upper limits could be set for those

elements not observed and these are given in table 46.

As can be seen from tables 44 and 45, some trace elements,

such as Ga, Ir, and the Cr dopant, appear to be fairly uniformly distributed within a given crystal. Others, especially Ba, may show differences of more than two orders of

83
Table 43. Results of activation analysis of ruby

crystals, group I, ppm.

Crystal Designation

Element 050(4) 050(9) 075(4) 075(9) 100(4) 100(9)

Cr

(#Cr ) 0.059 0.068 0.078 0.086 0.0104 0.098

Ga 0.024 0.030 0.070 0.096 0.034 0.039

Ir 0.040 0.037 0.033 0.038 0.023 0.033

Ba 89 111 68 5.3 1.1 16

Sr 2.5 3.0 1.9 <0.1 <0.1 0.5

Ca 0.051 0.089 0.050 0.054 0.038 0.028

Mn 0.0018 0.022 0.0023 0.0027 <0.001 <0.001

W 2.22 1.49 3.08 3.95 0.12 0.14

Au 0.0028 <0.0008 0.0014 <0.001 0.0021 <0.0003

La 0.036 0.027 0.011 <0.003 <0.004 <0.008

Table 44. Results of activation analysis of ruby

crystals, group II s ppm.

Cryst al "A" Cryst al "B" Crystal "C M Element 1 2 1 2

Cr

($Cr

2 3

) 0.0222 0.0215 0.0442 0.0463 0.0443

Ga 3.05 2.98 0.014 0.016 <0.004

Ir 0.239 0.245 0.050 0.054 0.026

Ba 42.6 0.24 15-2 1.26 9.4

Sr 1.55 <0.2 0.49 0.066 0.41

Cu 0.075 0.062 0.026 0.060 0.070

Mn 0.0013 0.002 - 0.0018 0.0012

W 0.032 0.052 0.015 0.0030 0.034

84
Table 45. Results of activation analysis of ruby

crystals, group III, ppm.

Crystal "D" Crystal "E" 12 12

Cr ($Cr

2 3

) 0.0373 0.0351 O.0389 0.0377

Ga

Ir

Ba

Sr

Cu

Mn

0.008 <0.006 0.030 0.031

0.045 0.041 0.052 0.051

0.28 1.74 16.5 265

<0.03 0.085 0.54 8.3

0.024 0.016 0.027 0.137

0.0008 0.0006 0.0009 0.056

magnitude between adjacent pieces of the same crystal. To

find out whether these extreme fluctuations could be caused

by Ba segregation within the crystals, two of these adjacent

pieces were analyzed for Ba by K. F. J. Heinrich and C. Piori

using electron probe microanalysis . In one of the pieces

they found actual inclusions which contained 25-30% Ba.

A paper describing this work was presented at the 159th

National Meeting of the American Chemical Society, Houston,

Texas, February 1970 and is now being prepared for publication
Table 46. Detection limits a for trace elements in

ruby, ppm.

Ag 0.05 Eu 0.0005 Mn 0.0005 Sr 0.1

Ar 0.1 Fe 5 Mo 0.05 Ta 0.1

As 0.002 Ga 0.003 Nd 1 *Tb 0.2

Au 0.0001 Gd 0.02 Ni 1 Te 0.6

Ba 0.1 *Ge 0.2 *0s 0.01 *Th 0.02

Br 0.01 Hf 0.001 Pd 10 *Tm 0.02

Cd 2 *Hg 0.5 Pr 0.05 *U 0.02

*Ce 0.1 Ho 0.2 *Pt 0.03 W 0.01

CI 0.05 I 0.01 Rb 1 Xe 0.01

Co 0.01 In 0.0001 *Re 0.001 Yb 0.01

Cr 1 Ir 0.005 Ru 1 Zn 0.5

Cs 0.5 K 1 Sb 0.02 Zr 100

Cu 0.005 Kr 1 Sc 0.05

Dy 0.001 La 0.003 Se 3

*Er 0.005 *Lu 0.02 Sm 0.001

'T'hese elements would be observed if present in ruby at or

above the levels indicated, assuming the following condi13 -2 -1 tions: 1 gm sample, 1 hour irradiation at 10 n . cm sec , p o 1 hour cooling time for decay of Al, 10 min counting time

for short-lived activities (<2 hr), 100 min counting time

for long-lived activities (>2 hr), all counting done with

^7-cc Ge(Li) detector. For elements marked * the limit will

be about a factor of ten poorer in the presence of Ir and

Cr at the levels normally observed to date. Elements not

listed would not be detected under these conditions even

if present.

(B. A. Thompson and E. C. Miller)

86
14

• Determination of Trace Elements in Synthetic Sapphire

Crystals

At the request of the Inorganic Materials Division two

pieces of a synthetic sapphire crystal were analyzed for

trace element content by neutron activation analysis. The

procedure followed was essentially the same as that previously

described for the analysis of ruby crystals. The crystals

were extremely pure and only a few elements could be observed.

Their concentrations are given in table 47.

Table 47. Trace elements observed in sapphire

crystals

.

Element Sample No. 1 Sample No. 2

Cu 0.027 ppm 0.038 ppm

Mn ^0.0005 ppm ^0.0005 ppm

Ir 0.002 0.0005 ppm

Sc ^0.015 ppm ^0.015 ppm

For the elements not observed the upper limits listed in

table 46 can be applied.

(B. A. Thompson)

5 . Determination of Gold in Semiconductor Silicon by

Neutron Activation Analysis

At the request of the Electronic Technology Division five

silicon wafers were analyzed for gold by neutron activation.

The most accurate possible values were required and therefore several special precautions were taken. The specimens

as received were broken approximately in half to obtain

samples which would fit horizontally in the irradiation container. Experiments by D. A. Becker (this report) have shown

there is no significant flux variation (<1%) within this

plane. Each sample (40-60 mg) was wrapped in polyethylene

and a small piece (^0.5 mg) of 1 mil gold foil was taped to

87
Table 48. Results of analysis of silicon wafers

Sample Designation

1250°C

1150°C

1050°C

950°C

850°C

Au Concentration, gm Au/gm Si

l.lO.xlO" 5

3.35xl0" 6

6.37x10

7.41x10'

2.89x10

-7

-7

the outside of the polyethylene to serve as a standard. The

samples were irradiated for 45 minutes in pneumatic tube RT4 of the NBSR which has the most highly thermalized flux of

the available facilities (cadmium ratio(Au) ^100). The

13 -2 -1

neutron flux was about 10 n.cm .sec . After irradiation

the samples were set aside for several days to allow any

short-lived activities to decay away. Each sample was rinsed

in HNO^ to remove surface contamination and then dissolved

in a small volume of HF-HNCU . Each solution was made up to

a standard volume and counted with a 47-cc Ge(Li) detector.

1 Q ft The 0.412 MeV peak of Au was used for analysis and no

y-rays corresponding to any other radionuclides were observed.

The gold foil standards were dissolved in aqua regia and a small aliquot of each was counted at the same geometry as the

samples. The results are shown in Table 48. The precision

of each value is estimated to be about ±2% (95% C.L.) and

the accuracy about ±5% (95% C.L.).

(B. A. Thompson)

P. The Nuclear Track Technique

1 . Introduction

The Nuclear Track Technique (NTT) is a method of analysis

in which tracks produced by charged particles in plastic

detectors are used for the determination of a few very specific elements. Usually the tracks are produced by either

alpha particles (e.g. for the analysis of boron and lithium)

88
or by fission fragments (e.g. uranium, and thorium). In

most cases, the element to be determined is placed in intimate contact with a suitable plastic detector, the sandwich irradiated with either neutrons or photons, the plastic

detector etched in a suitable etchant, and the track density

compared to a standard [30] or the weight fraction of the

element of interest calculated from known constants [31]

using the following equation.

C = fMp

k N R a At 6 Cos

o

r

f = multiplication factor (4 for boron, 1 for uranium)

M = Molecular weight of element determined — -2

p = Average track density (tracks cm ) N = Avogradro's number

o -2

<j)t = Thermal neutron fluence (neutrons, cm ) R = Range of fission or alpha particle (cm) in detector

o o c ~in a - Thermal neutron cross section for U(n,f) or B(n,a)

7 Li (cm)

6 = Density of material (grams, cm ) 2

Cos 6 = Etching efficiency factor; 6 is the cone half angle

at the tip of the etched track

K = Isotopic abundance of element determined

2 . Analysis of Uranium and Boron in Trace Elements in

Glass

a . Experimental - Uranium Analysis

The trace elements in glasses to be analyzed were received

in wafer form. They were selected cuts from 600 feet of glass

canes. The glasses were doped with trace elements at three

levels; 50 ppm, 1 ppm, and 0.02 ppm. Prior to analysis, the

glass wafers were cleaned by wiping the surfaces with alcohol

and then placing the wafers in a cold dilute nitric acid

solution (1 part HN0~ to 10 parts H„0) for 5 minutes. After
drying on a laminar flow clean bench, the wafers were weighed

and measured. The ^1 mm thick by ^13 mm diameter wafers were

placed on Lexan* polycarbonate slides, 20 mm wide x 40 mm

long x 0.25 mm thick. The sample and detector were then

bound together using paraffin wax. The resulting combination

was enclosed in polyethylene bags to prevent external contamination, and irradiated in the NBS reactor at a thermal neutron

13 -2 -1 flux of 1.1x10 n.cm .sec . The irradiation times were

dictated by expected concentration of uranium in the wafers.

After irradiation, the activity in the samples was allowed

24

to decay for two half lives of Na, since the majority of the

activity produced was from the ~Lk% Na

?

content of the glass

matrix.

After cooling, the plastic detectors were separated

from the wafers, and were etched for 45 minutes at 50°C in

Figure 23. Microscope display for fission track counting.

G.E. trade name for polycarbonate. See disclaimer in Preface.

90
6.5 N NaOH, rinsed in distilled water, and air dried. The

etched detectors were mounted on microscope slides with

tape. The tracks were counted with the use of a QTM quantimet* microscope. Using a magnification of 4800x on the

monitor, fission track counts within a prescribed area were

made in random fields of view (fig. 23).

During the course of analysis, however, it was discovered

235 that the U isotopic abundance in the glasses was different

from natural uranium, i.e. the isotopic abundance was 0.238%

for the 50 ppm glass, 0.280% for the 1 ppm glass, and 0.6l6%

for the 0.02 ppm glass. The analysis also indicated that

the uranium was homogeneously distributed throughout the

glasses. Typical results of the microscope fields observed

are given in tables 49, 50, and 51.

Table 49. Boron and uranium content in the trace

elements in glass (50 ppm).

Sample Boron Cone, (ppm) Uranium Cone, (ppm)

2A 31.22±1.49 a 36.90±0.78

37A 32.43±l-57 38.22±0.79

44A 32.51±2.95 36.45+0.78

68A 34.01±2.44 37.09±0.72

114A 31.78±1.47 36.02±0.87

Average 32.39±1.04 36.94±0.83

L ± values are the standard deviation of the mean (n=2)

Metals Research trade name for an image analyzing computer

microscope. See disclaimer in Preface.

91
b . Experimental - Boron Analysis

The boron analysis was similar to the uranium except

that cellulose acetate butyrate (CAB) detectors were used.

The CAB was used to record the alpha track produced from

10 7 nuclear reaction B(n,a) Li. These detectors were etched

after irradiation for eight minutes in 6.5 N HaOH at 70°C.

The boron distribution across the diameter of a wafer is

shown graphically in figure 24. The average number of

tracks across the diameter can be seen readily and is compared with the average value obtained from random fields

of view taken across the entire wafer.

50ppm BORON GLASS

Frame Size 4x|0" 6 cm2

Measure at 9.5x|0~6 C m

Intervals

Mean of values

10 20 30 40 50 60

FIELDS OF VIEW

Figure 24. Boron distribution across the diameter of a glass wafer.

(B. S. Carpenter'

92
Table 50. Boron and uranium content in the trace

elements in glass (1 ppm)

.

Sample

2A

36A

40A

69A

76A

114A

Boron Cone, (ppm)

1.293±0.154 a 1.251±0.201

1.379±0.135

1.287+0.086

1.296±0.135

1.227+0.160

Uranium Cone, (ppm ) 0.839+0.021

0.835+0.019

0.853±0.031

0.801+0.029

0.8l8±0.025

0.824±0.021

Average 1.295±0.052 0.828±0.01!

+ values are the standard deviation of the mean (n=2).

Table 51. Boron and uranium

elements in glass

content

(0.02 pp

in

m) . the trace

Sample Boron Cone. ( ppm) Ur anium Cone. (ppb

2A 0.199±0.033

a 72 .6±1.3

33A 0.208+0.033 72.1±1.2

43A 0.211+0.040 72.2±1.4

51A 0.20410.062 73-5+2.2

61A 0.206±0.058 72.0+1.1

76A 0.19510.037 - 130A 0.198+0.019 -

Average 0.203+0.057 72.5+0.2

l i values are the standard deviation of the mean (n=2).

93
3

Analysis of Orchard Leaves for Boron and Uranium

Previous work [30] had demonstrated that the analysis

of biological material by the NTT was possible. In order to

eliminate blank and standard problems present in the previously used methods, the direct determination method was

used. In this method prior to analysis the orchard leaves

were dried for 24 hours under vacuum at 90°C. Then 250 mg

of the finely powdered leaves were pressed into 13 mm diameter pellets which were 1.5 mm thick, using 20,000 lbs of

pressure from a hydraulic press.

The solid pellets were then bonded to plastic detectors

with paraffin wax for irradiation. CAB was used for boron

analysis and Lexan for uranium analysis. After irradiation

the slides were etched using the conditions discussed above,

and the average track density determined. The weight

fractions are then calculated from the track densities . At

present, the orchard leaves are still being analyzed.

(B. S. Carpenter)

4

Photo-fission Studies

Studies [32] have been continued on track production

from high energy photon induced fission of Bi, Te, Pb , Th,

and U. The first study entailed using different plastic

detectors to determine If the detection sensitivity varied.

Such detectors as Cronar* (a polyester), cellulose acetate,

cellulose acetate butyrate, mylar, and Lexan polycarbonate

were studied.

The second study was to determine the uranium weight

fraction to total track response using bremsstrahlung produced by 35 MeV electrons and a beam current of 20 uA . The

response was linear over the interval taken as is indicated

in figure 25.

Dupont trade name for polyester. See disclaimer in Preface

94
10

-7

LU

s lO"

<

8

<

OH

<

ZD

LUO

CO

<

10

-9

10

-10

10 10'

TRACKS OBSERVED

I0

V

Figure 25. Photon activation study depicting uranium

weight fraction to total track response.

Further work in this area will utilize both studies in

attempts to determine trace quantities of various elements

by photo-fission.

(B. S. Carpenter)

95
4. ACTIVATION ANALYSIS WITH THE COCKCROFT-WALTON

NEUTRON GENERATOR

A . Introduction

With a view to extending the usefulness of our neutron

generator laboratory, several research studies initiated

earlier [330 have been completed. These are; 1) Gamma-ray

spectra and experimental sensitivities for 3-MeV neutron

activation analysis; 2) 3-MeV neutron buildup interferences

in 14-MeV activation analysis; 3) Thermalization characteristics of 3-MeV neutrons and 4) Pulse pile-up and nuclear

interferences in 14-MeV neutron activation analysis for trace

oxygen. From an examination of the results reported herein

it is now possible to obtain a broad perspective on the applicability of this type of neutron source for the solution of

compositional problems in analytical chemistry. Typical

examples illustrating the capability of the neutron generator

as a versatile analytical tool are given. In this respect the

generator has been applied for the certification of several

organometallic Standard Reference Materials in the analysis

of which, thermalized neutrons, 3-MeV and 14-MeV neutrons

have been utilized.

In our continuing effort to establish a self-sufficient

fast neutron activation facility, we have developed techniques

and fabrication capability for reconditioning our own ion

sources. In the past the maintenance of this critical unit

involved a cost of several hundred dollars and a downtime of

at least one month. Currently it is possible to recondition

and replace key components of our Penning ion source and be

functional within a few hours at a cost of a few tens of

dollars. The ion source electrodes are fabricated in our

shops from materials whose suitability was investigated in

this laboratory.

As emphasized previously [33], o^v awareness of the health

physics problems associated with low-energy particle accelerators prompted a study of the possible neutron hazard

96
resulting from the bombardment of blank targets with a hydrogen ion beam. Although hydrogen-on-blank experiments are

routinely conducted during tune up operations , the possibility of a radiation hazard from neutron production by the

interaction of the deuterium component of hydrogen gas

with deuterons embedded in the target from previous irradiations, exists. Our results are indicative of the extent

of this hazard and provide guide lines for experimenters

in the safe operation of their individual facilities.

In conclusion, it is our opinion that the progress over

the past four years has promoted this laboratory to the

status of a self-sufficient and versatile fast neutron activation facility. The many investigations carried out and

reported in our earlier progress reports [33,34,35] over this

period of time have resulted in obtaining a sound analytical

competence which has been successfully employed not only for

the certification of standard materials, but to advance the

state of the art significantly. In the latter instance our

efforts in the achievement of high accuracy and precision

have been successful such that it is possible to evaluate

this technique in terms of other techniques of analysis

.

(S. S. Nargolwalla)

B . Research Activities

1 . Gamma-ray Spectra and Experimental Sensitivitie s for 3-MeV Neutron Activation Analysis

Past efforts in activation analysis with neutron generators have been primarily focused on the use of 14-MeV neutrons

because of the high yield from the D-T reaction. In certain

applications, however, 3-MeV neutrons from the D-D reaction

can be extremely useful. For example, interferences from

nuclear reactions having thresholds above 3 MeV are eliminated and the analysis f<~>r major components activated by 3-MeV

neutrons can be easily performed. In addition, some elements

have relatively high 3-MeV neutron activation cross sections

97
for (n,y) and (n,n T Y) reactions and are therefore about to be

determined in the presence of matrix materials which would be

highly activated by thermal or 14-MeV neutrons. Perhaps one of

the principal reasons for a comprehensive study on the 3-MeV

neutron activation of the elements stems from the fact that

3-MeV neutron buildup with time during 14-MeV neutron production could introduce significant systematic errors in 14-

MeV neutron activation analysis. Therefore if the sensitivities for 3-MeV neutron activation were known in conjunction

with the extent of 3-MeV neutron buildup with time on the

tritium target, it would be possible to estimate interferences from this mode of neutron production.

In this work, experimental sensitivities calculated from

gamma-ray emission from 73 elements irradiated with 3-MeV

neutrons have been determined. Critical information, such

Table 52. Experimental sensitivity for 3-MeV neutron

activation analysis

Sensitivity Range

Counts/g 10 n.cm sec Element

ca.

ca.

5

10

10

,6

6

>10^<10

4~ 5 >10 <\0 D

>10 <10

<ioNot Activatedb

Au

Ag

Cd

Sc

Cs

Re

Al

Rb

Gd

Be

Ca

Nb

Ho

Hg

W

In

V, Mn, Se, Sr, Y, Te, I,

Ba, Sm, Dy, Er, Lu, Ta,

Ir, Pt.

Cu, Zn, Ga, Ge, As, Br,

Mo, Ru, Rh, Pd, Sn, Eu,

Hf, Tl.

F, Na, Mg, Si, P, CI, K,

Ti, Cr, Fe, Co, Ni, Zr,

Sb, La, Ce, Pr, Nd, Tb

Tm, Yb, Pb

Li, C, N, 0, S, Bi

,10 min. irradiation, 0.5 min. decay and a 10 min. count.

Response less than twice background.

98
as detection geometry, gamma-ray counting efficiencies and

neutron source-sample configurations, has been provided to

facilitate application in other neutron generator laboratories. A summary of the results has been reported [36]. A

comprehensive paper describing the procedures in in press

[37]. A general compilation of the gamma-ray spectra obtained in this study is available from the authors. In table

52 the experimental sensitivities for 3-MeV neutron activation analysis have been summarized.

From table 52, analytical applications for 3-MeV neutron

activation can be predicted. Sensitivity ranges apply to

gamma energy peaks of greatest abundance.

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

2 . D-D Neutron Activation Interferences in 14-MeV

Neutron Activation Analysis

A source of systematic error in 14-MeV neutron activation

stems from the build-up of 3-MeV neutron flux. This buildup is due to neutron production from the interaction of the

primary deuteron beam with deuterons embedded in the tritium

target from previous bombardments. The extent of this buildup was studied using specially fabricated hydrogen-on-titanium

targets of various hydrogen loadings. The effects of various

parameters such as target loading and acceleration voltage

were studied. The 3-MeV neutron buildup as a function of time

at a constant beam current was measured with a neutron monitor

which had been previously calibrated as described earlier [33].

The results of the entire study are shown in figure 26 . It

may be noted that after only 1-2 mA hour of integrated beam

time the 3-MeV neutron flux extrapolated for a deuteron beam

r 1 current of 2.5 mA is of the order of 10 3-MeV n.cm sec

In fact this flux is about the same as when the best deuteron

targets were used in the sensitivity study described above.

From a knowledge of the build-up flux as a function of integrated beam time and 3-MeV neutron activation sensitivities

99
o

l CO

Figure 26

I l(T

CM

I

D-D NEUTRON BUILDUP VS. INTEGRATED BEAM

-D-D

sr==st^S3oo

x x-x—xD 6cm3

/in?H-in-Ti [200 kV, 1-OmA]

o 2crr#in? H-in-Ti [200 kV, 10mA]

X 8cm3

/in? H-in-Ti [200 kV, 10mA]

A 4cm3/in?H-in-Ti [!75kV, 10mA]

1.0 2.0

INTEGRATED BEAM [mA-HOUR]

3-MeV neutron buildup as

constant beam current.

a function of time at

(table 52), it is possible to estimate potential interferences in

any procedure using 14-MeV neutron activation. It is emphasized

that the specific data apply only to the facility being used.

However, the experimental concepts have been defined so that

similar data can be obtained in a relatively short period of time

for any other facility.

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

3 • Thermalization Characteristics of 3-MeV Neutron s Activation analysis with neutron generators has been

100
almost entirely limited to the use of 1^-MeV neutrons pro3 4 duced from the H(d,n) He reaction, although some use of

thermalized 14-MeV neutrons has also been made. The fact

2 3 that fluxes of 3-MeV neutrons from the H(d,n) He reaction

are about two orders of magnitude lower than the 14-MeV neutron

flux has discouraged the use of the lower energy neutrons.

Consequently, information in the literature regarding their

use for activation analysis is very limited.

A recent study [38] relating to the analysis of mixed

halogens in photographic emulsions suggested a detailed investigation into the applicability of 3-MeV neutrons for the

activation analysis of macroconstituents . In this report a summary of results obtained from a study of the thermalization characteristics of 3-MeV neutrons using "limited"

volume H„0 and D^O moderator is given. The experimental

arrangement was described previously [33].

Using bare and cadmium-covered indium foils, relative

flux measurements as a function of moderator thickness were

made. The flux profiles obtained were used to define optimum moderator thickness for both moderators. Using this

optimum thickness, measurements were made at the sample positions in our dual sample-biaxial rotating irradiation assembly

[39] using gold and yttrium samples for the determination of

slow (<0.l4 eV) and fast (>1 MeV) neutron fluxes respectively.

Under these conditions estimates of sensitivities for thermal

neutron irradiation can be computed.

Relative flux determinations as a function of moderator

thickness are illustrated in figure 27. At thicknesses of

approximately 6 and 8 cm, the thermal fluxes reach a maximum for HpO and D

? respectively. Absolute neutron fluxes

as measured at the sample position in our rotating irradiation device placed in contact with the moderator of optimum thickness were:

H o (6 cm.): 3-1x10 n.cm" 2

sec~

1 [<0.l4 eV]

h -P -1

3.6x10 n. cm sec [>1 MeV]

101
35

Q

LlI N

A.

i

i

i

i

i

i r H2 0[>0-l4eV]

H20[<0-l4eV]

N.B.

Errors are t.S for n = 3

VT\ [95%C.L.]

2 4 6 8 10

MODERATOR THICKNESS, cm

Figure 27. Relative flux determinations as a function of

moderator thickness.

D

2 (8 cm) 6.2xl0 3 n.cm

2

sec

X [<0.l4 eV]

2.1x10 n.cm~

2

sec

_1

[>1 MeV]

Without moderator, 3-MeV neutron fluxes at the sample position of our rotating device located at 6 and 8 cm from the

surface of the target cooling cap were 8.6x10 and 6.2x10

-2 -1 .. _ n.cm sec respectively.

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

102
4 . Pulse Pile-up and Nuclear Interferences in the Analysis for Trace Oxygen

In 14-MeV neutron activation analysis for trace oxygen,

nuclear interferences from most matrix elements can be eliminated if counting of the high energy gamma emission from the

0(n,p) N reaction is restricted to gamma energies greater

than about 4.5 MeV. The well known [40,4l] exceptions are

boron, fluorine and uranium. Oxygen determinations in these

matrices does not follow the normal comparative analytical

approach, and additional time-dependent matrix activation

information must be obtained. Further, pulse pile-up contributions from highly activated matrix elements are caused

by the inability of the normal scintillation detection system

to register individual gamma photons emitted at a high rate.

An earlier study [42] revealed pulse pile-up problems in the

analysis of trace oxygen in aluminum and copper. In this

report a comprehensive study of interferences for trace

oxygen determination in 22 different elements is described.

These elements serving as matrix materials were carefully

selected on the basis of their importance to industry, their

high 14-MeV neutron activation cross sections and unfavorable

decay schemes in terms of short half-life, emission of high

energy gamma photons and bremsstrahlung production.

The general method for the evaluation of these interferences consisted of simultaneously irradiating the sample

and standard in our rotating assembly, and sequential

counting between two 4-in. x 3- in. Nal(Tl) scintillation

detectors. Benzoic acid served both as an oxygen standard and

for flux normalization. Samples of the individual elements

were prepared to conform closely to the geometrical configuration of the benzoic acid standard. The experimental sequence controlled by our sequence programmer included a 30

second irradiation, 4 second decay, 22 second count (sample),

4 second decay and a 22 second count (standard). This sequence was repeated three times with a time interval of 2

103
minutes between sequences. The residual sample counts between the energy interval 4.8-8.0 MeV were accumulated in the

multiscaler mode of the pulse-height analyzer and used to

calculate the pulse pile-up contribution, and when possible,

the half-life of the interfering nuclide.

A summary of results to be published [43] is shown in

table 53. As expected the elements boron, fluorine and uranium

present severe problems even when oxygen is present at levels

of 1% or more. Interference from pulse pile-up is observed

in the case of aluminum and copper. Therefore, at oxygen

levels below 100 ppm in these elements, due consideration

must be given to this pulse pile-up effect. Expected interferences from the O.o second isomers Zr and Pb were

not observed due to the 4 second decay prior to counting the

sample. Finally, bremsstrahlung interferences resulting from

pQ pO

the Si(n,p) Al reaction was not observed because of the

2 cm thick lucite beta absorber employed in the design of our

detector system.

Table 53- Results of trace oxygen interferences.

Matrix Interference 51 ppm Equivalent Oxygen

Grams Matrix Element

Element Nuclear Pulse Pile-up

4 Boron 1.3x10

Fluorine 4.0x10^

Aluminum 10

Copper 20

4 Uranium 2.2x10

For the elements Be, N, Na, Si, P, Ti , V, Cr, Mn , Fe, Ni

Zn, Se, Zr, Mo, Ba and Pb , interferences, if any, were not

detectable

It is emphasized that the quantitative evaluations given

in table 53 apply to the particular conditions used in our

study. These conditions are normal in comparison to those

104
prevailing during our routine oxygen determinations at the

trace level. It is obvious that any significant alterations

of the experimental set up will alter the data. However,

the above investigation provides information which is useful

for designing suitable experiments for the determination of

trace oxygen in diverse matrices.

(S. S. Nargolwalla)

C . Analysis of Standard Reference Materials

In this area of analysis, the neutron generator was used

to provide comparative data in the case of two organo-metallic

SRMs . These were analyzed for their metallic component.

1

.

Cadmium Cyclohexane Butyrate SRM 1053a

Three samples drawn from the composite were analyzed.

Each sample weighed approximately one gram. Individual samples were irradiated simultaneously with a standard of SRM

1053 (Certified Cd=24%). Activation analysis of these samples

was performed using 3-MeV neutrons and counting the gamma

radiation from the reaction Cd(n,y) Cd and Cd(n, f Y)

lllm

Cd. A result of 23.35±0.25% Cd was obtained. The

uncertainty is that of the mean of 3 samples at the 95%

confidence level computed from ts//3, where s is the computed

standard deviation of a single sample with (3-1) degrees of

freedom.

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

2. Magnesium Cyclohexane Butyrate SRM 106lc

Six samples drawn from the bulk supply were analyzed.

Sample weights were about one gram each. Each sample was

irradiated with an equivalent weight standard (SRM 106lb; Mg=

6.53%). Activation of the samples were carried out with

14-MeV neutrons and counting the gamma radiation from the

Mg(n,p) Na reaction. A value for the magnesium concentration of 6.73±0.05% was obtained. The uncertainty in this

105
instance is of the mean of 6 samples at the 95% confidence

level computed from ts/Zo' , where s is the computed standard

deviation of a single sample with (6-1) degrees of freedom.

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

D . Special Analyses

1 . Oxygen in Metallic Barium

Two samples of metallic barium were received. Each sample

as received was encapsulated in a sealed 8 cm polyethylene

vial which was sealed inside a glass ampoule. The samples

were analyzed inside the polyethylene vials using the standard

14-MeV neutron comparative technique with benzoic acid as

the primary standard. Each sample was analyzed in duplicate

and the results obtained are as follows:

Capsule #1; 0.356±0.007% Oxygen

Capsule #2; . 358±0 . 007$ Oxygen

The errors indicated are weighted mean standard errors of the

weighted means based on Poisson counting statistics.

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

2 . Oxygen in Niobium Metal and Niobium-Titanium Alloy

Small samples of Nb metal and Nb-Ti alloy were analyzed

for oxygen using the standard technique for oxygen in metals.

Results for quadruplicate determinations obtained are given

below

Sample #35 (Nb); O.31±0.06f Oxygen

Sample #44 (Nb-Ti); 0.60±0.08# Oxygen

The errors quoted are the weighted mean standard errors of

the weighted mean of quadruplicate determinations based on

Poisson counting statistics. Corrections for all known

sources of systematic errors have been accounted for in the

final results.

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

106
E . Status of the Neutron Generator Facility

1

.

Reconditioning the Penning Ion Source

In the past it has been our experience that reconditioning and/or repair of our Penning ion source was both an

expensive and time consuming operation. We have therefore

developed the capability for reconditioning our ion sources

and can be operational within a few hours. In order to get

satisfactory results from our ion sources, experiments were

conducted to evaluate the life of several types of aluminum

as suitable materials for the Penning source detectors. Our

final conclusion was that both cathodes and anodes constructed

from either 2024 or 6061 aluminum served adequately. Pure

aluminum (type 1100) although possessing a high sputtering

threshold burnt out within a few minutes at high beam current

operation. It is expected that our reconditioned sources

will operate satisfactorily in excess of 200 hours at a beam

current of 2.5 mA. A factor of ten reduction both in terms

of cost and time has resulted from our own efforts to recondition our Penning ion source.

2

.

Health Physics

As mentioned in the introduction to this section, our concern In the practice of proper health physics precautions has

always been paramount. It is well known that neutron generator

operation manuals contain details for initial operation of the

generator using a hydrogen beam on a blank plate of tantalum or

copper. Our recent experiments concern the evaluation of a potential hazard from 3-MeV neutron emission. This hazard potentially exists since the small (^0.015% deuterium) amount of

deuterium contamination in hydrogen gas will generate 3-MeV neutrons by interacting with deuterium embedded in the target from

earlier bombardments. Two types of experiments were conducted.

In the first case hydrogen gas was allowed to bombard a deuterium loaded copper target, and in the second, hydrogen was allowed

to bombard a new blank copper target.

107
Initially a 2.5 mA deuterium beam was run on a clean

copper target for approximately 3 hours. As expected the

neutron flux built up steadily for the first 1 1/2 hours

and then leveled off. The steady state neutron flux was

5 -2 -1 measured to be approximately 1.4x10 n. cm sec at the sample

position in our rotating target assembly placed against the

generator target. The deuterium bottle was now replaced by

one containing hydrogen. The generator was then operated at

a beam current of 2.0 mA for 2 hours. The measured neutron

flux at saturation was about 1.8x10 n. cm sec . The generator was shut down for about 20 minutes. The gas lines

were evacuated and the vacuum pump allowed to pump down the

_ o entire system to 10 mm Hg.

The generator was then restarted and operated at 2.0 mA

beam current. The measured neutron flux after an irradiation

_2 of 2 hours on a clean copper target approximately lOOn.cm

sec" at the sample position. Since the maximum permissable

-2 -1

dose for 3-MeV neutron is of the order of lOn.cm sec the

potential hazard at the target position is about 10 times

greater. It is suggested as a general precaution that beam

viewing operations during tune up be extremely limited at

the target position.

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

108
5 . ACTIVATION ANALYSIS WITH THE NBS LINAC

A . Facilities

During the year, a new irradiation facility for the NBS

electron linear accelerator was designed and constructed.

Irradiation facilities at the NBS LINAC are used for a variety of purposes other than photon activation analysis.

These include the production of isotopes for Mossbauer and

radiochemical separations studies, measurements of photonuclear yields from thick target bremsstrahlung for nuclear

model interpretaiton, production of sources for decay scheme

elucidation and EM isotope separator experiments, and

measurement of high energy neutron and photofission yields

of the heavier elements. Irradiations are conducted with

electrons ranging in energy from 20 to 110 MeV. Planning

of a new facility required the accommodation of all of these

experiments, preferably permitting more than one irradiation

to be performed at the same time.

An irradiation facility consists essentially of three

components - a target made of high atomic number material

in which electrons radiate to produce bremsstrahlung, a sample transfer system to send samples from the laboratory

to the Irradiation position and return, and a beam dump or

backstop to absorb the intense photon beam. The irradiation

assembly is shown in figure 2 8.

1 . Target System

A major consideration in the planning of an irradiation

facility that is to accommodate a wide range of electron

energies is the radiation target configuration. In practice

one would desire a target that Is just thick enough to absorb the electron beam. Since, however, the target will be

irradiated by electrons of different energies, a compromise

is necessary.

The bremsstrahlung target consists of six water-cooled

1/16 in. tungsten foils. The foils are plated with a thin

109
Figure 28. LINAC irradiation assembly.

Figure 29. LINAC target assembly.

110
coating of gold to minimize corrosion of the tungsten. The

target is shown in figure 29.

Since it is difficult to weld tungsten to stainless

steel, the plates were press fit into the grooves provided

and the target holder welded shut. The one inch water inlet

line minimizes direct channeling of the water through the

center of the target. The front and back faces of the target

holder are 0.15 inch thick stainless steel to reduce power

deposition and subsequent heating by the electron beam.

This target will stop electrons of about hO MeV energies.

2

.

Sample Terminal

A dual irradiation terminal is located directly behind

the target which is loaded pneumatically from the laboratory

area in Service Area 2. One terminal is utilized for photon

energies of up to about 40 MeV. For higher energies a watercooled aluminum block eight inches thick is moved directly

behind the tungsten target to absorb unconverted electrons.

The samples are loaded end on in the terminal. The terminal

face is 15 mil thick stainless steel to minimize absorption

of photons

.

3

.

Beam Dump

Behind the terminal assembly is a 3^ in. x 3^ in. x 8 in.

stainless steel box which contains 52 two inch diameter lead

filled copper tubes (fig. 30). This beam dump absorbs the

intense photon beam and prevents it from striking structural

material or equipment. A high volume of water is circulated

around the tubes by a 1 1/2 in. supply system.

4

.

Pneumatic Transfer and Shuttle Systems

In order to effectively irradiate samples in the target

facility, a pneumatic transfer system was purchased from

Reactor Experiments, Inc. of San Carlos, California* (fig. 31)

The system consists of three separate air operated 7/8 in. dia-

*See disclaimer In Preface.

Ill
Figure 30. LINAC beam dump.

Figure 31. LINAC pneumatic transfer system control.

meter armoured polyethylene tubes. Two of these tubes are

equipped to send a sample into the terminal, sense photoelectrically when it is near the tip, and then brake the

samples' forward speed by the use of an air cushion arrangement. One of these two tubes will be connected to the low

energy irradiation terminal while the other will be used in

connection with a multi-position lazy suzan which will be

designed and built at a later date.

112
The third tube will be connected to the high energy

irradiation terminal and will be loaded in much the same

manner as the other two; however, at the discretion of the

operator the sample may be sent to Isotope Separator Building

instead of to the Irradiation terminal, thereby providing a shuttle system from the Photon Activation Laboratory in

Service Area #2 of the LINAC Building to the Isotope Separator Building nearly 800 ft. away. Transit times are expected

to be in the order of 10 sec. or less.

5. LINAC Interface System

In order that an electron beam can not be sent inadvertently to the photon activation target, a series of interlocks

must be closed. These interlocks are to insure that there

is cooling water flowing in all parts of the system and that

no unusually high temperatures exist at the critical points

.

The various temperature and flowrate interlocks are tied into

the accelerator by way of the LINAC Interface System (fig. 32)

It is also possible with this unit to monitor the flow rate

in G.P.M through the target and also the beam current through

the target.

Figure 32. LINAC interface system control

113
6 . Pneumatic Tube Coupled Lazy Susan

Plans have been made for a multi-position (10 or more) lazy

susan to be located in close proximity to the target. This

unit will have pneumatic tube capabilities and will be so constructed that samples can be inserted or removed at random.

(G. J. Lutz, F. A, Lundgren and J. E. Suddueth)

B . Analyses

1 . Determination of Thallium in Glass and Zinc Using

Photon Activation

a. Introduction

Neutron activation analysis has been used for the determination of thallium in various substances such as zinc and

zinc salts [44], potassium iodide [45], stone meteorites

[46] and rocks [47].

204 Thermal neutron activation leads to 3-75 year Tl and

O f\ fc. 4.3 min. Tl, neither of which is a gamma emitter. The

n a r- n a Z' low cross section (0.11 barn) for the reaction Tl(n,y) Tl

and the short half-life of the product make this reaction

less attractive for thallium determinations than the reaction

203 204 Tl(n,y) Tl. The thermal neutron cross section for the

latter reaction is about 11 barns.

The procedure described by Delberg et al. [45] utilizes

203 204 the reaction Tl(n,y) Tl with a sensitivity of approximately 0.1 ugm thallium. The determination of pure beta

emitters, however, requires rigorous chemical purifications

204 and the long half-life of Tl does not permit the observation of sample decay.

Past neutron activation has also been applied to the

determination of thallium [48], utilizing the reaction

203 Tl(n,2n) 202Tl. Thallium 202 has a convenient half-life

of 12.4 days and has a 0.44 MeV gamma ray. The threshold

for this reaction is in the order of 8-9 MeV and a relatively

poor sensitivity of about 20 ugm is obtained for an irradiflu:

114

13 2 ation of 30 hours at a total flux of 10 n/cm /sec.
Thus the fact that thermal neutron activation produces

pure beta emitters of thallium and fast neutron activation

yields poor sensitivities make the determination of thallium

by photonuclear activation most attractive. The work described here was undertaken to evaluate the potential for

determining trace amounts of thallium in various matrices by

2 en 202 the induction of the reaction Tl(y,n) Tl with bremsstrahlung from an electron linear accelerator.

b . Chemistry of Thallium

Thallium has two stable oxidation states, univalent and

trivalent, whose chemical properties differ markedly. Thallium (I) ions resemble the heavy alkali ions and lead and

silver ions while thallium (III) ions are similar to indium

and gallium ions. Thallium (III) is a strong oxidizing agent

and forms complexes with various inorganic and organic liquids

The existance of two stable oxidation states constitutes the

basis of a number of analytical methods. Various combinations

of thallium (III) extraction with ethyl ether and isopropyl

ether, Tl (I) extractions with N-benzoyl-N-phenyl-hydroxylamine (BPHA) [49], and Tl (I) precipitations of Iodide and

chromate have been used to obtain purified thallium in neutron

activation analysis. A procedure has also been described for

the separation of thallium (I) by amalgam exchange [50].

It was observed during the course of an investigation

of a chelating agent that thallium (III) and gold (III) extracted quantitatively into benzyl alcohol and octyl alcohol

from various concentrations of acid solution. Mercury (II)

gave a partial extraction under these conditions (on the

order of 35% extracted per extraction). Thallium (I) did

not extract into either alcohol.

These extractions were applied to the chemical separation

of thallium after irradiation by bremsstrahlung from the

NBS electron linear accelerator in NBS standard glass samples

and various zinc salts and zinc metal.

115
c

Irradiation Facility

Most irradiations were performed in the M5 facility of

the NBS LINAC . The electron energy was about 35 MeV with

an average beam current of about ^OuA. Bremsstrahlung was

produced by allowing the electron beam to radiate in a water

cooled tungsten target. The converter, target and pneumatic

transfer system have been described previously [51]. The

samples were usually irradiated 1-2 hours

An auxiliary irradiation facility has been placed in the

magnet room of the NBS LINAC. The bremsstrahlung is obtained from extraneous scatter from the slit in another

experimenters facility. The flux obtained in this facility

is lower by a factor of about 10 than the facility previously described. The electron energy used in this facility is typically about 100 MeV. Although the higher electron

energy makes the possibility of interfering reactions more

probable, we feel much useful data has been obtained using

this facility. Samples are transported from the laboratory

to the irradiation facility and returned using a rabbit

system which uses compressed air to facilitate transfer.

d. Standards and Flux Monitors

Samples of thallium sulfate were encapsulated in polyethylene snap cap vials of the same size as those used for

the samples and were used as standards . Accurately weighed

discs of copper were taped to each end of the vials as flux

monitors and the vials were placed in screw cap aluminum

rabbits. The TlpSCv standard was treated chemically in the

same manner as the particular sample of interest.

eExperimental Procedure for the Determination of

Thallium in Glass Samples

Samples weighing approximately two grams were encapsulated in polyethylene snap cap vials and copper disc flux

monitors were placed on each end of the vial. After encapsulation in a screw cap aluminum rabbit the samples were irradi116
ated in the 45° facility of the NBS LINAC for 1-3 hours

depending on the level of thallium in the sample.

The glass was dissolved in 10 ml perchloric acid, with

a few ml of HF and approximately 20 mg TlpSO^ carrier added.

Thallium (I) was oxidized to thallium (III) by the addition

of a small amount of potassium chlorate solution. The mixture was diluted to 30 ml with distilled H

? 0, cooled to room

temperature, and extracted twice with 15 ml portions of octyl

alcohol. The combined organic phases were washed three times

with 6N HC1 and the separator funnels centrifuged each time

to facilitate the separation of layers. The organic phase

was collected in a 2 oz. polyethylene bottle and counted

using a 47cc Ge(Li) detector. The T^SO^ standard was treated

and extracted under similar conditions. The results obtained

are reported in table 54.

Table 54. Results of analyses for thallium in NBS glass.

Sample Number Thallium (ppm)

95GCG SRM 610 66., 66., 62., 6l., 6l.

95GCE SRM 612 15 . , 15., 12.

95G SRM 6l4 0.20, 0.24

The use of a Ge(Li) detector permits the determination

of thallium without a separation of gold. However, if desired, the gold can be readily separated before the extraction step by adding Au(III) carrier and passing S0 ? through

the solution to reduce gold to the metal and filtering.

Thallium is reduced to thallium (I) during this step but may

readily be oxidized again to thallium (III) and then extracted.

f . Experimental Procedure for the Determination of

Thallium in Zinc and Zinc Salts

Samples weighing 1-3 grams were encapsulated in snap cap

polyethylene vials and copper discs taped to each end to

serve as flux monitors. The activation was performed in the

117
auxiliary facility using another experimenters spray at a maximum electron energy of approximately 100 MeV. Since the

flux in this facility was much lower than that obtained in

the 45° facility irradiation times of up to 8 hours were used.

The activated sample was dissolved in HC1 and approxmately 20 mg thallium carrier added. Thallium (I) was oxidized to thallium (III) by the addition of H„0„ and the

sample heated to expel excess peroxide. The solution was

diluted with distilled water to give a final HC1 concentration of about 1M . The sample was cooled to room temperature

and extracted twice with 15 ml portions of benzyl alcohol

that had been pre-equilibriated with IN HC1. The separatory

funnels were centrifuged at ^1000 rpm for 5 minutes to facilitate separation of layers and the organic phase washed at

least five times with IN HC1 with added zinc and copper carrier.

Some zinc which extracts under these conditions , is sometimes

not completely washed from the organic phase. In these cases

it is necessary to allow decay of the Zn which has a 440

202 keV gamma ray - the same as that of Tl. The half-life of

69mr7 . n

i. , Zn is 14 hours

The organic phase was counted on a 47cc Ge(Li) detector

to obtain the thallium activity. The results obtained are

presented in table 55-

Table 55. Thallium content in zinc and zinc salts.

Sample ppm Thallium

Zinc Metal 1.5, 2.0

Zinc Chloride 2.5, 2.5, 3.5

Zinc Acetate 7-0, 8.0, 5.4, 6.7

Octyl alcohol can also be used to extract the thallium

in this procedure but the HC1 concentration must be approximately 6N. Using octyl alcohol the thallium is not quantitatively extracted until the HC1 concentration is ^6N. But

at this concentration a large amount of zinc is also extracted

118
and problems were encountered in removing all the zinc by

washings

.

The spread in the results in table 55 can be attributed mainly to poor counting statistics. A typical sample

of 7 ppm thallium if counted in the Ge(Li) detector for 1200

min produces only 621 counts in the peak channels . The background for the same number of channels and for the same time

is 468 counts. Thus, the net counts attributable to the

sample in this instance have a standard deviation due to

counting of approximately 21$

.

The sensitivity of this method should be increased significantly with the installation of a new target assembly

that is in progress at the present time.

g. Interfering Reaction

A possible interference in the photon activation pro20 4 20 2 cedure for thallium is the nuclear reaction Pb(y,2n) Pb

.

202 202 The Pb then decays by electron capture to Tl.

A sample of high purity Pb metal was irradiated in the

45° facility at 35 MeV for two hours to evaluate the possible

Pb interference. The lead was dissolved, thallium carrier

added, and thallium oxidized to the thallium (III) state

with chlorate. The solution was extracted twice with 15 ml

benzyl alcohol and the combined organic phases washed three

times with IN HNO^. The organic layer was counted in a 47cc Ge(Li) detector and compared to a T1„S0k standard that

had been extracted into benzyl alcohol and counted under

similar conditions. The sum of the counts in the thallium

channels was not significantly more than the background indicated that Interference from this source is negligible at

35 MeV.

(L. Masters and G. J. Lutz)

119
2. Determination of Gold, Tantalum, Manganese and Iron

in Cadmium Metal by Photon Activation

a. Introduction

Samples of one-fourth inch cadmium rod were received in

the Section from the Office of Standard Reference Materials

for determination of gold, tantalum, manganese and iron.

Because of the extreme difficulties that would be encountered

doing an analysis of cadmium by thermal neutron activation

analysis due to neutron self-shielding, it was decided to

accomplish the determinations by photon activation analysis.

b . Experimental Determination of Au and Ta

The cadmium metal rod weighing approximately 4-5 gms was

irradiated in the NBS LINAC . Tantalum foil and gold foil

were used as standards and were positioned so they surrounded

the cadmium metal. The tantalum foil standard was dissolved

in HF and the gold foil In aqua regia for comparison with the

sample

The cadmium metal was dissolved in nitric acid, diluted

to about 6N HNO„, and approximately 10 mg Au carrier was added

The solution was extracted twice with 15 ml benzyl alcohol

that was pre-equilibrated with 6N HN0~. The separatory

funnels were centrifuged for a few minutes at ^1000 rpm to

facilitate the separation of layers. The organic phases

were combined and washed 4-5 times with 6N HNCi containing

approximately 10 mg/ml cadmium. The aqueous phase and the

washings were kept for the Ta analysis

The combined organic phases were drained into a 2 oz.

polyethylene bottle and counted for Au using a 47cc Ge(Li)

detector. An aliquot of the gold standard was also extracted

into benzyl alcohol and counted under the same conditions as

the sample

The combined aqueous phase and the washings were made to

2M HF and extracted twice with 15 ml methyl isobutyl ketone

(MIBK) . The tantalum is quantitatively extracted into MIBK

120
under these conditions. The combined organic phases were

washed several times with a mixture of HNO--HF of the same

concentration as the aqueous phase.

The organic phase was drained into a 2 oz . polyethylene

bottle and counted on a 47cc Ge(Li) detector for Ta.

This tantalum isotope has a half-life of 8.1 hours so the

tantalum portion was counted before the gold portion. The

tantalum standard was extracted and counted under similar

conditions

.

c . Determination of Mn

The cadmium metal rod weighing approximately 4-5 gms

was irradiated in the NBS LINAC . Manganese metal chips were

placed around the sample and were used as the standard.

After irradiation the cadmium metal was dissolved in

6N HNO and the solution heated. Approximately 100 mg Mn

carrier followed by sodium bromate were added. The manganese

was precipitated under these conditions as MnO ?

. The precipitate was collected on filter paper using a Buchner

apparatus and washed several times with 6N HN0„. The MnO^

was then dissolved in 3N HC1 and silver nitrate added to precipitate AgCl. The mixture was filtered and the precipitate

discarded. This precipitation was repeated several times to

remove the various silver isotopes formed from stable cadmium

isotopes during the irradiation. The solution and the

washings were combined and evaporated to 30 ml and this

solution counted on the 47cc Ge(Li) detector. An HC1

solution of the manganese standard was counted under conditions similar to those of the sample. The sum of the counts

in the channels of the 0.84 MeV Mn peak were used to

quantitate the results.

b . Determination of Fe

The cadmium metal rod surrounded by chips of the Fe foil

57 56

was irradiated and the reaction Fe(y»p) Mn was used for the

determination of iron.

121
The sample was dissolved in 6N HNC> and the solution heated.

Approximately 100 mg Mn carrier followed by sodium bromate

were added. The manganese was precipitated as MnO„ and the

precipitate collected by filtration. The MnOp was dissolved

in HC1 and AgCl precipitations as in the procedure for manganese . The resulting solution was evaporated as rapidly as possible to ^30 ml and the 0.8^5 MeV peak of 2.58 hr J Mn used

to quantitate the manganese. The iron standard was dissolved

In HNCu and the manganese precipitated as MnO using NaBrC> . The precipitate was dissolved in HC1 and the solution counted

56 under the same conditions as the sample for Mn activity.

(G. J. Lutz and L. Masters)

3- Determination of Nitrogen in Metals by Photon

Activation

Preliminary experiments are being conducted to develop

reliable techniques for the determination of nitrogen in

metals by photon activation analysis. The reaction induced

14 13 is N(y,n) N. The product nucleus decays by positron emission and has a half-life of 10.0 minutes.

Generally, the most difficult problem in developing

methods for determination of the light elements by photon

activation is assuring a clean quantitative separation. It

appears that for many metals, after irradiation and dissolution in an oxidizing medium, the nitrogen is present as

ammonium ion. This can be liberated as ammonia by addition

of concentrated sodium hydroxide. It is difficult to distill

all of the radioammonia quantitatively in the relatively

short period of time required so a known amount of ammonia

was added before the sodium hydroxide. The sodium hydroxide

is added to the solution in a closed glass system and the

solution is heated and agitated for several minutes . Radioammonia is trapped in a known amount of sulfuric acid.

13 After counting the annihilation radiation from the N by

122
coincidence spectrometry with 2 Nal detectors, the sulfuric

acid is back-titrated to determine the yield. Yields of

60-80$ are typical.

The standards are ammonium sulfate. After irradiation,

the standards are placed in a special aluminum counting container of sufficient thickness to annihilate all of the

positrons in a small volume. They were counted with one

detector only at a distance of about eighteen inches. The

counting container was then filled with copper powder in

64 which Cu, a positron emitter, had been induced. This was

counted at exactly the same geometry as the nitrogen standards

The copper powder was then dissolved in nitric acid and an

aliquot of the solution counted under the same conditions as

64 the sample. The ratio of the counting rate of the Cu

allowed a computation of the counting correction factor between samples and nitrogen standards.

Fluxes are monitored with copper. The reaction Cu

£-\ o C^ o no

(y,n) Cu is induced. Cu has the same half-life as N

and is thus very satisfactory for this purpose.

A few runs have been made on some NBS Standard Reference

Materials already certified for nitrogen. These were in the

form of chips. The results for ferrous alloys were within

^15% of the certified values, which is moderately acceptable

at this preliminary stage of our investigations. Alloys of

titanium and zirconium gave unreasonably high values. This

is attributed to recoil nitrogen in the air. The chips are

too small to etch. Inert atomsphere facilities for packaging will be required.

(G. J. Lutz and L. Masters)

123
6. APPLICATION OF COMPUTERS TO ACTIVATION ANALYSIS

A. I ntroduction

Assistance from computers can be of great benefit in

data handling and computation of results for activation

analysis, thereby relieving the experimenter of much tedious

work. The application of computer techniques to achieving

this goal is described in this section of the progress report

Before proceeding with the description of what has been

accomplished, it may be valuable to present an overview of

some of the problems in data handling and result computation,

and to indicate how these problems are partially or completely

solved by computer techniques.

Computers are useful in two distinct, but not unrelated,

areas of activation analysis data processing. These areas,

alluded to above, may be labeled data acquisition and

manipulation (DAM), and result computation (RC) or computation of results (COR) . The first area (DAM) could also be

called data handling, but this often implies result computation also. Because the two areas are actually quite distinct, they should have distinct labels. Therefore, what we

shall call data acquisition and manipulation refers first of

all to retrieval of data from experimental apparatus; for

example, the recording on magnetic tape of a spectrum from a multichannel analyzer is one type of retrieval. This step in

process is the acquisition of the data.

Once the data has been acquired, hopefully in a form

compatible with the computer to be used, it may be stored

for use at a later date, or it may be processed by the computer for result computation. It is likely that the data,

as originally recorded, will need further manipulation before

the result computation can begin. This manipulation involves

translation of the data into appropriate form for use internally by the computer. Additional manipulation may be

required since the Initial data recording may contain errors,

124
and these errors must be taken into account before processing

the data for result computation. Manipulation may also be

performed to create permanent records of certain data, such

as catalog spectra. The term data acquisition and manipulation, therefore, includes readout of the data from the

analytical instrument, storage of the data, correction of

errors in the record, translation into internal computer format, and permanent file generation. After completion of the

data acquisition and minipulation phase, the data has been

stored in the core of the computer as real or integer numbers.

Result computation then proceeds by the location of peaks,

computation of peak areas, calculation of concentration (e.g.

parts per million), performance of decay curve analysis, and

so on. Some of the advantages of digital computers for data

acquisition and manipulation will become apparent from the

following example. Suppose an analyst takes a spectrum using

a Ge(Li) detector and a 4096 channel analyzer. If the memory

capacity is six decimal digits (i.e. 0-999999) then each

experiment produces 6x4096 or 24576 digits of data. If lead

zeros are suppressed, the number of digits to be considered

is significantly reduced, but, even so, a teletype printout

of 4096 channels of data occupies 512 printed lines and about

0.75 hours of readout time while a paper printer will give

4096 lines; clearly, this is a very cumbersome amount of

data to record, store, or even scan visually. If the spectrum

is a complicated one, it may have as many as 50 peaks, and it

will be a tedious, time consuming chore to locate the peaks

in which the analyst Is interested. It is, therefore, very

advantageous to let a computer help with the analysis . The

computer can easily cope with the 24576 digits, assuming the

analyzer has furnished the data in computer compatible form.

Once the spectrum is In core, the computer can quickly

search for peaks, calculate peak areas and print the peak

locations and areas.

125
Having briefly indicated some of the advantages of

computer processing of activation analysis data, it is

pertinent to discuss some of the problems.

The first problem is acquisition in computer format of

the information accumulated by the radiation measuring device.

If an appreciable volume of data is accumulated, i.e. by a multichannel analyzer, readout onto magnetic tape is often

available. These readouts are not necessarily in a convenient format for utilization in the computer. Records

written on magnetic tape by multichannel analyzers are

usually too long to be handled by FORTRAN, since the inputoutput buffer is arranged to handle one line of print, 132

characters. Two of the analyzers used by the Activation

Analysis Section write the data in blocks of 15^2 characters

(256 channels of data plus one tag word, 6 characters per

word or channel). Fortunately, the UNIVAC 1108 software

subroutine, NTRAN, makes it easy to read the longer tape

records; it should be pointed out, however, that previous

experience with an IBM-36O was very different, for it took

several months, several hundred dollars, and several man

weeks to read a similar tape with our own assembly language

program. Reading a block of data with NTRAN gives an

array of the UNIVAC equivalent of BCD characters. These

characters must then be converted into integer or real

numbers. Illegal characters on the tape can be a problem

here .Once the data are in core and ready for some sort of

numeric processing, we have to contend with statistical

scatter in the data and with the fact that no mathematical

function exists which describes more than a small region of

a Ge(Li) detector spectrum. The numerical analysis of gamma

ray spectra has received considerable attention from many

scientists, and fairly good progress has been made [52]. Considerable progress is needed in estimation of minor components

126
in Ge(Li) detector spectra and in decay curve analysis,

however

.

In particular, the computation of net full energy peak

areas is a very important one for activation analysis, since

many activation analysis determinations utilize peak areas

obtained from Ge(Li) or Nal(Tl) spectra. The problem lies in

the calculation of base area, which is subtracted from the

gross area, to obtain the desired net peak area. If the

net area is large, relative to the base area, there is no

problem because a significant error in the base area will not

result in a significant relative error In the net area.

Very often the opposite situation occurs, however, and we

need to base quantitative results on peaks which have small

net areas compared to base areas. In this case, shape and

height of the baseline are crucial. Some methods of studying this problem have been published [53, 5*0, but further

study is still needed.

Other topics which need study are complex peak resolution, decay curve analysis, and component identification.

An example of interfering peaks which would benefit from

numerical resolution is the interference between the 559 keV

As peak and the 564 keV Sb peak. These Isotopes are

frequently found together, especially in the analysis of

steels

.

Our group sometimes analyzes samples about which little

is known concerning chemical composition. It would be ver"y

useful if we had a computer routine to examine the Ge(Li)

detector spectra and Identify the spectral components.

Particular attention must be given, in this type of analysis,

to those components which are obscured by other, more intense

components, and to those components which are present at

amounts just above their detectable limits.

In the foregoing paragraphs a brief summary of areas of

application of computer techniques to activation analysis

has been given. Some of the problem cases have been outlined

127
roughly. In the remainder of this section the progress that

has been made since this work was begun last October (1969)

is discussed. . . (H. P. Yule)

B. ALSPIS - Activation Analysis Peak Area Program

ALSPIS is a Fortran V computer program which locates

peaks in Ge(Li) detector spectra, and computes their areas,

widths, and energies. Data are read from cards or from

magnetic tape, and checked for errors. Various options

allow the analyst considerable control over the course of

the subsequent calculations. Details of the program are

indicated below. The program is by no means finished, for

it will have to be changed to meet the changing requirements

of the activation analysis group, and the basic algorithm

can be modified, improved, extended, or expanded with new

ideas; or a new alogrithm can be written.

ALSPIS (pronounced: all-spice) may be considered a two part program with the addition of a third part to control

program flow. Data are read into the computer and manipulated by part one, and result computation is performed in

part two. The three parts will be described below in order.

The data manipulation may be best described In terms of

its development. At the outset, our group had a TMC-4096

channel analyzer coupled to a DATAMEC* magnetic tape unit.

All that was known about the form of the data on the tape

other than the density, 200 BPI , was that a four digit tag

word was written on the tape. Operation of the units had

been limited to using the tape for storing spectra which

could later be read back into the analyzer.

Inspection of the tape unit indicated that it probably

wrote BCD characters on the tape, and using the utility

routine, TPRINT, available at the NBS UNIVAC-1108, it was

found that each channel of the data (data word) is represented by six BCD characters and that each record on the

*See disclaimer In Preface.

128
tape consists of an unbroken string of 257 words. The first

two characters of the first x^ord in a record are the address

of the 256 channel memory segment in the analyzer, and the

next four are the tag word which the analyst set at the time

of recording. The remaining 256 words are data. If more

than 256 channels of data are written on the tape, the

analyzer records them in groups (records) of 257 words as

described above - a 2048 channel spectrum contains eight

such groups, for example. The parity of the recording is

even.

As indicated in the introduction, it is not possible to

read the 15^2-character records with a FORTRAN READ statement

since it is limited to about 132 characters. However, the

utility subroutine NTRAN* may be used for this purpose, and

the 15^2-character record is stored in an array of 257 words

of 6 characters each. The next step is to convert them from

a string of characters to real numbers.

An easy way to accomplish this is to write them onto a magnetic drum scratch area, using a FORTRAN write statement

with an alphanumeric format, rewind the drum, and read the

data back Into core using a FORTRAN statement with a numeric

format. The program simply repeats the process as many times

as necessary to read a complete spectrum.

Data manipulation in this way is sufficient to provide

numbers for result computations unless errors occurred when

the tape was written. These errors are of two possible types.

The first type is the transcription of a non-numeric character

by improperly functioning electronics, and the second is

transcription of records which are not the prescribed length

of 15^2 characters. The second type of error is easily

overcome by programming to ignore records not having 15^2

characters . Most of the records so ignored are short and

^Unless specifically stated otherwise, program names used

are utility routines available at the NBS Computer Center.

129
meaningless. Reading of the data from the drum is not

possible when there are errors of the first type, and at the

time this part of the data manipulation was being programmed,

an attempt to read a non-numeric character on a numeric

format resulted in termination of the reading process and

proceeding to the next executable statement. Thus, to overcome errors of type one, the invalid character must first

be located and then altered to be acceptable to the computer

(if not the analyst).

One way to locate the bad character is to reread the

1542 characters into a 1542 word array, and compare them

against the information contained in table 56. The digits

zero through nine, the alphabet, and some punctuation marks

are listed in column 1. Each of these characters is represented internally in the 1108, by a particular 36-bit string.

The octal representation of the bit string for each character

is shown in column 2, and the same "number" printed out on

an integer (base ten) format is shown in column 3- Inspection

of table 56 shows that valid numeric characters must be in

the base ten range -17094651578 to -7430975162 (inclusive).

Characters outside this range will not be the digits 0-9,

and thus it is easy to detect non-numeric or good characters

without testing each one to determine if it is one of the ten

acceptable digits. Using this technique requires care to

avoid overflows. An IF statement comparing a letter "Z" and

a numeric "0" might result in the subtraction of the "0" from

the "Z":

i.e. 33371214149

+17094651578

50465865727

while the computer gets, incorrectly, -18253611008. To

avoid this difficulty, the first test must determine whether

the character is "positive" (invalid) or "negative" (valid or

invalid), and the second test, on "negative" characters

determines If the value Is in the range indicated above.

130
Table 56. Representation of alphanumeric character

string internally and as base ten numbers

600505050505 -17094651 57;-

1 610505050505 -16020909754

2 620505050505 -1494716793H

3 630505050505 -1 387342o.l0o

U 640505050505 -1 279968428 2 5 650505050505 -1172594P45S

6 660505050505 -1 f,65220nb34

7 670505050505 -957845*81 '< 8 700505050505 -850471b98^

9 710505050505 -743097R16?

A 060505050505 b52766A54^

6 070505050505 760l41.i373

C 100505050505 867515P197

D 110505050505 9748P.94021

F 120505050505 1 08228.^845

F 130505050505 1189637768?

G 140505050505 1297011949^

H 150505050505 11043861317

I 160505050505 1511760314.1

J 170505050505 lol9l34496=>

K 200505050505 1726508^789

L 210505050505 18336828613

M 220505050505 19412570437

N 230505050505 2i4863lP2hi

240505050505 2156005U085

P 250505050505 2263379^909

Q 260505050505 23707537733

R 270505050505 24781279557

S 300505050505 25855021381

T 310505050505 2o92876^205

U 320505050505 2800250502°

V 330505050505 29076248853

W 340505050505 30149988677

X 350505050505 31223730501

Y 360505050505 32297472325

Z 370505050505 33371214149

r 560505050505 -192421.35225

750505050505 -3136007866

+ 420505050505 -32127037114

/ 740505050505 -M209749690

S 470505050505 -267583?799'i

* 500505050505 -25684588170

( 510505050505 -2461084434b

) 400505050505 -3427452r,7b2

131
After a bad character is detected, it must be converted to

some numeric character and the repaired array written on a drum so that the reading on a numeric format may continue.

When an error is discovered, the magnitude of the data word

(channel) in which the bad character(s) occurred is set to

zero. It would be possible in some cases to estimate the

number of counts In the bad channel if it occurred singly,

but errors frequently have appeared in groups of successive

channels, and the remedy of setting the bad channels to

zero and printing out their channel numbers has been sufficient . After the error correction part of ALSPIS was completed

and debugged, the read statement which has the form

READ(i, j ,ERR=k, END=1) list

i = unit number

j = format statement number

k = a statement number

1 = a statement number

list= the variables being set by the reading process

was modified by the NBS Computer Services Department.

Originally, the clause, ERR=k, caused a transfer to statement number k in the event an error occurred In data transmission in the computer. The modification resulted in

invoking the transfer whenever an attempt was made to read

invalid data, such as attempting to read non-numeric characters

on a numeric format. This is a very convenient feature, and

the Computer Services Department Is to be commended. It Is

unfortunate that computer manufacturers did not implement

this feature long ago.

Error locating and correction using this feature would be

somewhat simpler than the method described above, but it

is pointless to replace that method since it is debugged and

operational and errors occur Infrequently.

132
After the spectrum, with any corrections, has been stored

in the computer core by the data manipulation part of ALSPIS,

the result computation phase, part two of ALSPIS, is initiated. The Ge(Li) detector spectrum is searched for significant peaks, and their net and gross, boundary channels, full

widths at half maximum, energies, and the standard deviations

of the net areas are calculated and are printed. A significant peak is arbitrarily chosen as having a net area of more

than 50 counts and a standard deviation of 25 percent or less.

The analyst has the freedom to modify these criteria. The

basic method of peak location utilizes spectrum smoothing

and differentiation, which has been published and is summerized in reference [52]. Advantages of the method are speed

and sensitivity. Computation time for a 4096 channel spectrum

is a few seconds or less, and peaks which are quite small,

relative to background, are as easily located as more prominent

ones . The major problem in computing net peak areas is in

calculation of the underlying base area. The problem is

trivial if the net area is large relative to the base area

and there are no interfering peaks or Compton edges . The

top of the base Is well approximated by a straight line

tangent to the spectrum at either side of the peak. On the

other hand, the top of the base area is probably not accurately

given by a straight line when the net area is small or is

small relative to the base area. The selection of the junctions of the line representing the base top with the spectrum

itself may be quite difficult when an accurate net area is

sought. Clearly, computation of the base area is difficult

even for prominent peaks if they are not cleanly resolved

from other proximate peaks

.

These problems are accounted for in the peak area computations of ALSPIS without requiring appreciable amounts of

computer time, and the analyst is provided with information

on all peaks in the spectrum. By inspection of the results

133
printed, he can determine how well (i.e. with what relative

error) a net area is determined, and accept the result (which

often happens), or use his own calculation method (which is

often not really much better) if the peak is not well resolved

or is small relative to its base. The latter choice is indicative of the need for further study to improve ALSPIS.

In operation, the peaks and boundary channels are located

by finding sign changes in the smoothed first derivative [55,

56]. The net area is computed as indicated in ref. [55], and

the results are printed. Boundary channels for the same peak

are then re-evaluated by searching for minima in the smoothed

spectrum. Using these boundary channels, a second net area

for the same peak is computed and printed. Of course, the

second area is different from the first only if one or both

boundary channels change as a result of the re-evaluation.

A second attempt at boundary modification uses the slope

method [52], yielding an area computed with a third set of

boundary channels. A fourth evaluation involves the regularity of the behavior of the smoothed first derivative; specifically the derivatives should approach zero monotomically as

the channel number changes from half height toward the boundary channel. If it does not, the possible presence of an

interfering peak is suggested. This behavior of the derivative supplies a fourth area for the same peak. A fifth way

to determine peak boundary channel is to use the smoothed

first derivative calculated via a different formula [57, 58]

which also tends to detect interfering peaks. The last

(sixth) method of peak area computation uses smoothed counts

in the boundary channels for the base area computation.

Thus, areas are computed with boundary channels selected

by six different methods. Usually the same boundaries are

selected by several of the methods. If each of the six

methods yields substantially the same net area, it is very

likely that the peak is free of interference and its net

area is accurately calculated for use in subsequent compu134
tations. If poor agreement is obtained, then there is

either an interfering peak or the peak area is subject to

uncertainties because of its small size relative to its

base. In this case, the analyst should examine the results,

and he is then able to accept one of the results, compute

the area himself, use a different peak from the same isotope,

or use a more elaborate computer program for that particular

peak. It should be noted that the accuracy of the general

peak area computation has been established [56].

Part three of ALSPIS provides progress flow, guided by

data control cards supplied by the analyst. These cards

inform the program of various options such as the source of

the data (magnetic tape or cards), and various input parameters such as energy coefficients, peak size limits, and

so forth. Considerable allowance is made for improperly

punched cards, and many spectra may be processed with only

a few data control cards. With all parameters set, either

by the program or the analyst, the data manipulation phase

begins, and when complete, result computation begins. Depending on information specified by the data control cards, computation is followed by data manipulation of a new spectrum,

reading of additional data control cards, or job termination.

(H. P. Yule)

C . A Mini-Computer and the Activation Analysis Section

The Activation Analysis Section recently acquired a used

Varian-622/i* minicomputer with an 18 bit word size and an

8192 word memory (fig. 33). Basically, there are two ways in

which we might choose to use such a computer: (1) for data

acquisition and manipulation and (2) for various computations

including result computation. There are many ways In which

to implement these choices, especially with the addition of

various peripherals such as the magnetic drum, not delivered

until some time after the computer arrived, and various

*See disclaimer in Preface.

135
Figure 33. Varian 622/i minicomputer.

magnetic tape units. One possible way to use the computer

would be to have it act as a secretary, acquiring spectral

data, and storing it on magnetic tape for subsequent processing by the UNIVAC 1108. The data could also be relayed via

telephone lines for remote batch processing. Alternately,

the Varian could be used to perform the result computations

as well as to acquire, manipulate, and store it. In this

event, the printing speed of the teletype becomes a factor,

136
and it would be possible to use the UNIVAC 1108 as a result

printer, transmitting results to it either by telephone line

or magnetic tape.

With these and many other analyzer-computer-peripheral

configurations to be considered, the decision on the optional

arrangement clearly could not be made without having a thorough knowledge of the capabilities of the Varian. A

relatively near term goal, which could be sought prior to

deciding on the final configuration, is to have the computer

acquire data directly from the multichannel analyzers or

other devices, and transcribe the data on magnetic tape for

storage and processing by the UNIVAC.

What has been done, therefore, is to work with the Varian

to acquire a thorough knowledge of its capabilities. In this

way it has been possible to make steady progress toward

several goals: (1) the near term analyzer to magnetic tape

capability, (2) a decision on the final configuration of the

analyzer-computer-peripheral system, (3) determination of

its capabilities as a computer including hardware and software, and (4) acquisition of a knowledge of its operation

also in terms of both hardware and software.

The first step in this course of action was to attend a two week course, given by Varian, to learn to operate the

computer and to learn to program in the 622/i machine language and also the assembly language, DAS. The next two

months were spent trying to get the machine running, since

it was not operable when received and had many hardware bugs.

During this period, the magnetic drum arrived, but both hardware and some of the software were of the 16 bit word variety,

(Varian also manufactures the similar 620/i with 16 bit words)

and Varian had to modify the drum controller board and send

for new software from the factory. Debugged, the drum has

a capability of storing 27,648 words, 18 bits per word.

During this period, the TYMSHARE 620/i simulator was used,

but it was not too helpful because of differences between

137
the TYMSHARE 620/i computers (16 bit words, 12K word memory)

and our machine.

It was apparent from watching machine operation in class

that the speed of the teletype (10 characters per second) as

an input/output device would severly limit the rate of job

throughput. In particular, the reading of punched paper

tapes containing source or binary programs consumed considerable time. Therefore a high speed paper tape reader (300

characters per second) was added, effectively reducing

program compilation time by roughly a factor of two. In

some instances the factor of 30 could be realized - loading

of the DAS assembler was reduced from roughly half an hour

to one minute, for example. The reader was fully operational

when installed, but no supporting software was supplied, and

it was necessary to modify the paper tape loader program and

the DAS assembler to read paper tapes with the high speed

reader rather than the teletype.

With the system operational, a number of programs have

been written to approach the goals indicated above. It is

not necessary to list all these programs here, and it will

be sufficient to indicate briefly the current state of

sophistication. The DAS assembler, utility program, and

an executive program are stored in the drum, and these

programs are easily reloaded in the event of a disasterous

programming error. The program to do this is called EXEC-I;

it utilizes the Varian supplied drum driver and a number of

routines written here. Binary to decimal and decimal to

octal subroutines have been written since the Varian supplied

software for these purposes is not adequate to handle our

requirements . (Their binary to decimal routine is limited

to non-negative numbers less than 10000, for example). Gamma

ray spectra may be read from punched paper tape, converted

to binary, and stored on the drum. This spectrum may be

smoothed or differentiated. Peaks may be located using the

smoothed first derivative, and peak areas computed using

138
double precision integer arithmetic.

The next step will be to program the computer to interrupt

computation, save registers and core using the drum, accept

a spectrum from an analyzer, store it, and return to the

original computation. In this regard, the speed of the teletype for printing results is beginning to be a severe limitation . Further future developments will seek the goals indicated above. It is also clear at this point that a FORTRAN

Compiler is needed. , TT „ „ n

. (H. P. Yule)

D- Miscellaneous

A number of existing computer programs were modified

and tested for use by our section. These include CLSQ, the

decay curve analysis program [59] 3 and SAMPO, the complex

photopeak resolution program [53]. A study of upper limits

to non-resolved components was initiated with Prof. Emile

Schweikert of the Activation Analysis Research Laboratory

at Texas A&M University. A number of programs were written

for our section to use with the on-line computers

.

(H. P. Yule)

139
7. PERSONNEL AND ACTIVITIES

A . Personnel Listing

Activation Analysis Section

Philip D. LaFleur, Section Chief

Betty E. Crowther, Secretary

Patricia Case (Student Aid)

Activation Analysis, Reactor

P. D. LaFleur, Project Leader

T. E. Gills

W. D. Kinard

H. L. Rook

J. S. Olin, Guest Worker, Smithsonian Institution

Stephanie Snead (Summer Worker)

Activation Analysis, Cockcroft-Walton Generator

S. S. Nargolwalla, Project Leader and Assistant

Section Chief

J. E. Suddueth

Sheryl Birkhead (Summer Worker)

G. W. Smith (Transferred)

Activation Analysis, LINAC

G. J. Lutz, Project Leader (1)

L. W. Masters (1/2)

Radiochemical Separations

B. A. Thompson, Project Leader

W. F. Mar low, Guest Worker, USAEC

L. W. Masters (1/2)

Activation Analysis of Biological Materials

D. A. Becker, Project Leader

Nuclear Data Handling

H. P. Yule, Project Leader

Nuclear Track Technique

B. S. Carpenter, Project Leader

140
B. Publications

P. D. LaFleur, "Determination of Molybdenum in Steels and

Biological Materials by Activation Analysis," Radiochem.

Radioanal. Letters, 1, 255, (1969).

L. T. McClendon, I. H. Qureshi, and P. D. LaPleur,

"Extraction Studies of the Group IIIB-VIIB Elements and the

Lanthanides Utilizing Bis ( 2-Ethyl-Hexyl ) Orthophosphoric Acid,"

Radiochim. Acta, 12, 107, (1969).

B. A. Thompson and P. D. LaFleur, "Activation Analysis

for Molybdenum in Samples Containing Large Amounts of Tungsten,"

Anal. Chem. 4l, 1888, (1969).

G. J. Lutz and P. D. LaFleur, "Determination of Yttrium

in Rare Earths by Photon Activation Analysis," Talanta, 16 , 1457, (1970).

G. J. Lutz, "Determination of Oxygen in Sodium by Photon

Activation Analysis," Anal. Chem. 42_, 531, (1970).

G. J. Lutz, "Determination of Carbon in High Purity

Metals by Photon Activation Analysis," Anal. Chem. 4_2, 948,

(1970).

S. S. Nargolwalla, M. R. Crambes and J. E. Suddueth,

"Photon Self-Absorption Corrections for the Minimization of

Systematic Errors in 14-MeV Neutron Activation Analysis,"

Anal. Chim. Acta, 49., 402, (1970).

B. S. Carpenter and C. H. Cheek, "Trace Determination of

Uranium in Biological Material by Fission Track Counting,"

Anal. Chem. 4_2, 121, (1970).

G. J. Lutz, "Forensic Science: A Bibliography of Activation Analysis Papers," NBS Tech. Note 519 (1970).

G. J. Lutz, "Determination of the Light Elements in Metals:

A Bibliography of Activation Analysis Papers," NBS Tech. Note

524 (1970).

G. J. Lutz, "Pollution Analysis: A Bibliography of

the Literature of Activation Analysis," NBS Tech. Note 532

(1970).

G. J. Lutz, "14-MeV Neutron Generators in Activation

Analysis: A Bibliography," NBS Tech. Note 533 (1970).

G. J. Lutz, "Oceanography: A Bibliography of Selected

Activation Analysis Literature," NBS Tech. Note 534 (1970).

141
"Register of Personnel Working in Activation Analysis"

NBS Miscellaneous publication.

P. D. LaFleur, "Laboratory of the Issue," J. of Radioanal.

Chem. 3, 127 (1969).

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

"Experimental Sensitivities for 3-MeV Neutron Activation

Analysis," J. Radioanal. Chem. (in press).

D. A. Becker and P. D. LaFleur, "Production and Certification of NBS Biological Standard Reference Materials,"

Proceedings, 4th Annual Conference on Trace Substances in

Environmental Health, June 1970, Columbia, Mo. (in press).

S. S. Nargolwalla, "Application of Neutron Generators

to Activation Analysis," Proceedings, Conference on Applications of Small Accelerators, Oak Ridge, Tennessee, March

1970 (in press)

C. List of Talks

P. D. LaFleur, "Determination of Cerium, Molybdenum and

Scandium in Steel and Rock Samples Using a Radiochemical

Separation and Selective Stripping," 158th National Meeting

American Chemical Society, New York, New York, 8 September

1969.

B. A. Thompson, "The Determination of Trace Elements in

Glass by Activation Analysis Using Hydrated Antimony Pentoxide

for Sodium Removal," (T. E. Gills, W. F. Marlow, and B. A.

Thompson) 158th National Meeting, American Chemical Society,

New York, New York, 8 September 1969.

D. A. Becker, "Neutron Activation Analysis for Uranium

Using a Rapid Radiochemical Separation," (D. A. Becker and

P. D. LaFleur) 158th National Meeting, American Chemical

Society, New York, New York, 8 September 1969-

G. J. Lutz, "The Determination of Oxygen in Sodium by

Photon Activation Analysis," 158th National Meeting, American

Chemical Society, New York, New York, 8 September 1969.

L. W. Masters, "The Extraction of the Transition Elements

Using the Antibiotic, Tetracycline," 158th National Meeting,

American Chemical Society, New York, New York, 12 September

1969.

D. A. Becker, "Neutron Activation Analysis for Sodium at

the Picogram Level," American Nuclear Society, Winter Meeting,

San Francisco, 2 December 1969

142
T. E. Gills, "The Determination of Short-Live Trace

Impurities in Sodium Metal," (T. E. Gills and"" P . D. LaPleur)

American Nuclear Society, Winter Meeting, San Francisco,

2 December, 1969.

B. A. Thompson, "Analysis of Medeival Stained Glass by

Neutron Activation," (J. S. Olin, E. C. Miller and B. A.

Thompson) American Nuclear Society, Winter Meeting, San

Francisco, 2 December 1969 .

P. D. LaFleur, "Novel Techniques in Neutron Activation

Analysis," American Nuclear Society, Winter Meeting, San

Francisco, 3 December 1969.

S. S. Nargolwalla, "Experimental Sensitivities for 3-MeV

Neutron Activation Analysis," (S. S. Nargolwalla and J. E.

Suddueth) American Nuclear Society, Winter Meeting, San

Francisco, 4 December 1969

.

G. J. Lutz, "Determination of Carbon in Metals by Photon

Activation Analysis," (G. J. Lutz and L. W. Masters) American

Nuclear Society, Winter Meeting, 4 December 1969.

H. L. Rook, "The Determination of Oxygen in Silicon by

Charged Particle Activation Analysis," (H. L. Rook and E. A.

Schweikert, Texas A&M University) American Chemical Society

Meeting, Houston, Texas, 24 February 1970.

S. S. Nargolwalla, "Advances in Neutron Generator Activation Analysis," American Chemical Society Meeting, Houston,

Texas, 24 February 1970.

B. A. Thompson, "Determination of Trace Impurities in

Ruby Laser Crystals by Activation Analysis," American Chemical

Society Meeting, Houston, Texas, 24 February 1970.

D. A. Becker, "Characterization of a Nuclear Reactor for

Neutron Activation Analysis," (D. A. Becker and P. D. LaFleur)

American Chemical Society Meeting, Houston, Texas, 24 February

1970.

G. J. Lutz, "Photonuclear Activation Analysis," American

Chemical Society Meeting, Houston, Texas, 24 February 1970.

P. D. LaFleur, "Radiochemical Separations In Activation

Analysis," American Chemical Society Meeting, 24 February

1970.

P. D. LaFleur, "Trace Analysis: A Frontier of Analytical

Chemistry," Seton Hall University 1969-1970 Guest Seminar

Program of Eminent Scientists, South Orange, New Jersey,

17 February 1970.

143
S. S. Nargolwalla, "Application of Neutron Generators to

Activation Analysis," Conference on Small Accelerators for

Teaching and Research, Oak Ridge Associated Universities,

Oak Ridge, Tennessee, 24 March 1970.

B. S. Carpenter, "The Nuclear Track Technique, A Manual

Means of Activation Analysis," Morgan State College, Baltimore

1 May 1970.

P. D. LaFleur, "Activation Analysis," Renssalaer PolyTechnic Institute, Troy, New York, 25 May 1970.

P. D. LaFleur, "Standards for Activation Analysis: Are

There Any?" Joint Chemical Institute of Canada and American

Chemical Society Conference, Toronto, Canada, 27 May 1970.

B. S. Carpenter, "The Nuclear Track Technique: A Growing

Microanalytical Tool," Joint Chemical Institute of Canada and

American Chemical Society Conference, Toronto, Canada, 28 May

1970.

G. J. Lutz, "Determination of Carbon and Oxygen in Sodium

by Photon Activation Analysis," Society of Applied Spectroscopy, 21st Mid-America Symposium on Spectroscopy, Chicago,

3 June 1970.

D. A. Becker, "Production and Certification of NBS Biological Standard Reference Materials," (D. A. Becker and

P. D. LaFleur) 4th Annual Conference on Trace Substances in

Environmental Health, University of Missouri, Columbia,

Missouri, 25 June 1970.

B. S. Carpenter, (paper read by L. A. Currie) "Boron and

Uranium Determination in Simulated Lunar Glass by the Nuclear

Track Technique," Annual Meeting of the American Nuclear

Society, Los Angeles, 1 July 1970.

S. S. Nargolwalla, (paper read by L. A. Currie) Determination of Pulse Pile-up and Nuclear Interferences in 14-MeV

Neutron Activation Analysis for Trace Oxygen." Annual Meeting

of the American Nuclear Society, Los Angeles, 1 July 1970.

1^4
8. ACKNOWLEDGMENTS

The Activation Analysis Section wishes to thank Dr.

Robert S. Carter, Mr. Tawfik M. Raby, Mr. James P. 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 wish to thank the U. S. Atomic Energy Commission for

making possible the visit of Mr. W. P. Marlow.

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 have made our entire operation more efficient and

effective

.

145
9. REFERENCES

[1] LaFleur, P. D. Editor, NBS Tech. Note 508_, 1*1 (1969).

[2] ASTM Publication E 262-65T, "Measuring Thermal Neutron

Flux by Radioactivation Techniques," published by ASTM,

Philadelphia, Pa. (1965).

[3] DeVoe, J. R., Editor, NBS Tech. Note 40_4, 14, (1966).

[4] LaFleur, P. D., Editor, NBS Tech. Note 5C_8, 11, (1969).

[5] Euler, B. A., Covell, D. F. and Yamamoto, S., Nucl.

Inst, and Meth. 7_2, 143 (1969).

[6] Olin, J. S. and Sayre, E. V., Proceedings. Modern Trends

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Vol. 1, U. S. Government Printing Office, Washington,

D. C, (1969).

Perlman, I. and Asaro, F., Archeometry 11, 21 (1969).

LaFleur, P. D., Editor, NBS Tech. Note 5C_8, 34, (1969).

Coleman, R. F. and Wood, G. A., AWRE Report No. 03/68

(April 1968)

Turner, W. E. S., J. Soc. Glass Technology 40_, 39

(1956); ibid. 40, 162 (1956); ibid. 4p_, 277 (1956).

"On Divers Arts," The Treatise of Theophilus, translated

by J. G. Hawthorne and C. S. Smith, University of

Chicago Press, Chicago, 111., (1963).

[12] Osmond, R. G. and Smales, A. A., Anal. Chim. Acta 10 , 117 (1954).

[13] Knight, J. D., Novey, T. B., Cannon, C. V., and

Turkevich, A., Radiochemical Studies: The Fission

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1916 (195D.

[14] Debrun, J., Barrandon, J., and Albert, Ph., Proceedings,

Modern Trends in Activation Analysis, NBS Special Publication 312, Vol. II (1969).

[15] Friedlander, G., Kennedy, J., and Miller, J., Nuclear

and Radiochemistry , John Wiley and Sons, New York, 95

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146

[7]

[8]

[9]

[10]

[11]
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(1968).

[18] Girardi, P. and Sabbioni, E., J. of Radioanal. Chem. 1_, 1969 (1968).

[19] Smith, G. W. and Becker, D. A., "Preparation of an NBS

Biological Standard Reference Material for Trace Elemental Analysis," Proceedings, Nuclear Activation Techniques in the Life Sciences, IAEA (1967).

[20] Certificate of Analysis [Provisional], Standard Reference Material 685, High Purity Gold, U. S. Dept . of

Commerce, National Bureau of Standards.

[21] DeVoe, J. R. , Editor, NBS Tech. Note 276, 72 (1965).

[22] Becker, D. A. and Smith, G. W., Proceedings, Modern

Trends in Activation Analysis, College Station, Texas,

230 (1965).

[23] Thompson, B. A. and LaFleur, P. D., Anal. Chem. 4l_, 852

(1969).

[24] Thompson, B. A. and LaFleur, P. D. Anal. Chem. 4l , 1889 (1969).

[25] Thompson, B. A., Proceedings, Modern Trends in Activation Analysis, NBS Special Publication 312, Vol. 1,

636 (1968).

[26] LaFleur, P. D., Editor, NBS Tech. Note 5^8, 40 (1969).

[27] LaFleur, P. D., Editor, NBS Tech. Note 458, 41

(1968).

[28] Covell, D. F., Anal. Chem. 3J_, 1785 (1959).

[29] DeVoe, J. R. Editor, NBS Tech. Note 404, 42 (1966).

[30] Carpenter, B. S. and Cheek, C. H., Anal. Chem. 42_, 121

(1970).

[31] Fleischer, R. L. and Price, P. B., Science, 144 , 84l

(1964).

[32] Taylor, J. K., Editor, NBS Tech. Note 50_5, 110

(1969).

147
[33] LaFleur, P. D., Editor, NBS Tech. Note 508., 65

(1969).

[34] LaFleur, P. D., Editor, NBS Tech. Note 458 , 9

(1968).

[351 DeVoe, J. R., Editor, NBS Tech. Note 4_28, 13 (1967).

[36] Nargolwalla, S. S., Trans. Am. Nucl. Soc . 12 (2), 510

(1969).

[37] Nargolwalla, S. S. Niewodniczanski, J. and Suddueth,

J. E.j J. Radioanal. Chem. (in press).

[38] Przyby lowicz , E. P., Smith, G. W., Suddueth, J. E., and

Nargolwalla, S. S., Anal. Chem. 4l, 819 (1969).

[39] Lundgren, F. A. and Nargolwalla, S. S., Anal. Chem. 4_0_, 672 (1968).

[40] Anders, 0. U. and Brichen, D. W., Anal. Chem. 36, 287

(1965)

[41] Wood, D. E., Tech. Note TN-105, Kaman Nuclear, Colorado

Springs, Colorado, June 1965.

[42] Hoste, J., DeSoete, D., and Speecke, A., EUR-3565E,

University of Ghent, Ghent, Belgium, June 1967.

[43] Nargolwalla, S. S., Suddueth, J. E., and Rook, H. L.,

Trans. Am. Nucl. Soc. (in press).

[44] Gibbons, D. and Lawson, D., Proceedings, Modern Trends

in Activation Analysis, College Station, Texas, 172,

(1965).

[45] Delbecq, C. J., Glendenin, L. E., and Yuster, P. H.,

Anal. Chem. 2J5, 350 (1953).

[46] Ehmann, W. D. and Huizenga, J. R., Geochem. Cosmochim

Acta 17, 125 (1959)

[47] Morris, D. F. C. and Killick, R. A., Talanta 4, 51

(I960).

[48] Talat-Erbin, M. and Okan, S., C.N.A.E.M. - 39, Cekmece

Nuclear Research Center, Istanbul, Turkey, Sep. 1966.

[49] Lyle, S. J. and Shedrikar, A. D., Anal. Chim. Acta 32,

575 (1965).

148
[50] Qureshi, I. H. , and Meinke, W. W., Radiochim. Acta. 2,

99 (1963).

[51] DeVoe, J. R., Editor, NBS Tech. Note 428 , 72 (1967).

[52] See, for example, H. P. Yule, "Computation of Experimental Results in Activation Analysis," in J. R. DeVoe,

Editor, Modern Trends in Activation Analysis, NBS

Special Publication 312, Vol II, 1155 (1969).

[53] Routti, J. T. and Prussin, S. G., UCRL-I7672.

[54] Quittner, P., Anal. Chem. 4l, 1504 (1969).

[55] Yule, H. P., Anal. Chem. _38, 103 (1966).

[56] Yule, H. P., Anal. Chem 40_, 1480 (1968).

[57] Heydorn, K., private communication, (1968).

[58] Barnes, V., IEEE Trans. Nucl. Soc. NS-15 , 437 (1968).

[59] Cumming, J. B., BNL-6470, also NAS-NS 3107, 25 (1963).

149

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