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Q NBS SPECIAL PUBLICATION 528

U.S. DEPARTMENT OF COMMERCE/ National Bureau of Standards
NATIONAL BUREAU OF STANDARDS

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

Bureau's overall goal is to strengthen and advance the Nation's science and technology and

facilitate their effective application for public benefit. To this end, the Bureau conducts

research and provides: (1) a basis for the Nation's physical measurement system, (2) scientific and technological services for industry and government, (3) a technical basis for equity in trade, and (4) technical services to promote public safety. The Bureau's technical work is performed by the National Measurement Laboratory, the National Engineering Laboratory,

and the Institute for Computer Sciences and Technology.

THE NATIONAL MEASUREMENT LABORATORY provides the national system of

physical and chemical and materials measurement; coordinates the system with measurement

systems of other nations and furnishes essential services leading to accurate and uniform

physical and chemical measurement throughout the Nation's scientific community, industry,

and commerce; conducts materials research leading to improved methods of measurement,

standards, and data on the properties of materials needed by industry, commerce, educational

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Agencies; develops, produces, and distributes Standard Reference Materials; and provides

calibration services. The Laboratory consists of the following centers:

Absolute Physical Quantities2 — Radiation Research — Thermodynamics and

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services to users in the public and private sectors to address national needs and to solve national problems in the public interest; conducts research in engineering and applied science

in support of objectives in these efforts; builds and maintains competence in the necessary

disciplines required to carry out this research and technical service; develops engineering data and measurement capabilities; provides engineering measurement traceability services; develops test methods and proposes engineering standards and code changes; develops and

proposes new engineering practices; and develops and improves mechanisms to transfer

results of its research to the utlimate user. The Laboratory consists of the following centers:

Applied Mathematics — Electronics and Electrical Engineering2 — Mechanical

Engineering and Process Technology2 — Building Technology — Fire Research —

Consumer Product Technology — Field Methods.

THE INSTITUTE FOR COMPUTER SCIENCES AND TECHNOLOGY conducts

research and provides scientific and technical services to aid Federal Agencies in the selection, acquisition, application, and use of computer technology to improve effectiveness and

economy in Government operations in accordance with Public Law 89-306 (40 U.S.C. 759), relevant Executive Orders, and other directives; carries out this mission by managing the Federal Information Processing Standards Program, developing Federal ADP standards

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provides the technical foundation for computer-related policies of the Federal Government.

The Institute consists of the following divisions:

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'Headquarters and Laboratories at Gaithersburg, Maryland, unless otherwise noted; mailing address Washington, D.C. 20234.

: Some divisions within the center are located at Boulder, Colorado, 80303.

The National Bureau of Standards was reorganized, effective April 9, 1978.
NOV 2 8 1978

Analytical Methods for Safeguards and Accountability

Measurements of Special Nuclear Materials

Proceedings from American Nuclear Society

Topical Meeting held May 15-17, 1978

Williamsburg, Virginia

Edited by

H. Thomas Yolken

National Measurement Laboratory

National Bureau of Standards

Washington, D.C. 20234

JohnE. Bullard

Babcock & Wilcox Company

Lynchburg Research Center

Lynchburg, Virginia 24505

Sponsored by

Nuclear Fuel Cycle Division

Virginia Section American Nuclear Society and

National Measurement Laboratory

National Bureau of Standards

Washington, D.C. 20234

U.S. DEPARTMENT OF COMMERCE, Juanita M. Kreps, Secretary

Dr. Sidney Harman, Under Secretary

Jordan J. Baruch, Assistant Secretary for Science and Technology

NATIONAL BUREAU OF STANDARDS, Ernest Ambler, Director

Issued November 1978

t
Identification of commercial products does

not imply recommendation or endorsement by

the National Bureau of Standards nor that the

product is necessarily the best available for the

purpose.

Library of Congress Catalog Card Number: 78-6001 15

National Bureau of Standards Special Publication 528

Nat. Bur. Stand. (U.S.), Spec. Publ. 528, 297 pages (Nov. 1978) CODEN: XNBSAV

U.S. GOVERNMENT PRINTING OFFICE

WASHINGTON: 1978

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

Stock No. OO3-003-O1996-8 Price $4.75

(Add 25 percent additional for other than U.S. mailing).
CONFERENCE OFFICERS

GENERAL CHAIRMAN

T. C. Engelder (B&W)

ASSOCIATE GENERAL CHAIRMAN

R. L. Hoffman (B&W)

TECHNICAL PROGRAM COMMITTEE

Pickles, Chairman (LLL)

J. Busch (NBL)

M. Delmastro (Allied Chemical)

L. Hoffman (B&W)

C. Whitehead (NBL)

T. Yolken (NBS)

GUEST PROGRAM

Judy Connell (B&W)

FINANCE

William Bishop (B&W)

HOTEL LIAISON

Evan Morgan (B&W)

REGISTRATION

Tom Williamson (UVA)

PUBLICATIONS

John E. Bullard (B&W)

REGISTRATION COMMITTEE

Tom Williamson (UVA)

D. R. Tulodieski (B&W)

J. D. Patterson (NBS)

Edited by

H. Thomas Yolken and John E. Bullard

ACKNOWLEDGMENT

The editors are especially appreciative of the outstanding help

provided by Ms. Dianne Patterson on the administrative aspects of

the publication.
ABSTRACT

This book contains proceedings of the American Nuclear Society's

(co-sponsored by the National Bureau of Standards) topical meeting

entitled, "Analytical Methods for Safeguards and Accountability

Measurements of Special Nuclear Materials. " The meeting was held

in Williamsburg, Virginia, on May 15-17, 1978. The two objectives

of the meeting were to discuss the latest techniques for chemical

analysis of special nuclear materials and to strengthen lines of

communication among scientists working in this field.

The presentations deal with advanced analytical chemistry

techniques such as x-ray spectrometry for elemental content, neutron

interrogation and gamma ray techniques for isotopic content, mathematical correlation models, and wet chemistry methods for elemental

content. Examples of remote systems for handling highly radioactive

samples for analysis are given. Progress is being made on the problems

associated with obtaining high precision and accuracy for analysis

of special nuclear materials while at the same time maintaining short

turn around time and reduced possibility of diversion. Various

approaches are described which emphasize computer-controlled operating

systems with built-in safeguards and quality assurance programs.

Current trends in the fuel cycle, along with future approaches to control and accountability of special nuclear materials, are discussed.

KEYWORDS: Nuclear safeguards; accountability; x-ray; gamma-ray;

analysis; special nuclear materials; neutron

iv
CONTENTS

essian I - May 15, 1978; 8:45 a.m. Chairperson - R. L, Hoffman

Babcock & Wilcox

Opening Remarks - T. C. Engelder

Babcock & Wilcox

The Chemical and Isotopic Analysis of Uranium,

Plutonium and Thorium in Nuclear Fuel Materials

C. E. Pietri, J. S. Palter and C. D. Bingham (Invited)

New Brunswick Laboratory, U.S. Department of Energy,

Argonne, Illinois

PAGE

1

Analytical Methods and Laboratories for Safeguards

and Accountability: Present and Future

L. T. Corbin (Invited)

Oak Ridge National Laboratory, Oak Ridge, Tennessee 19

An Automated X-ray Spectrometer for Mixed Oxide Pellets

M. C. Lambert (Invited) 3 M. V. Goheen, M. W. Urie, and N. Wynhoff

Hanford Engineering Development Laboratory, Richland, Washington 20

Mechanisms for Assuring Accurate Measurement of Nuclear Materials

H. T. Yolken (Invited)

National Bureau of Standards, Washington, D.C 37

Design of a Cf-252 Neutron Assay System for Evaluation

at the Savannah River Plant Fuel Fabrication Facility

T. W. Crane, S. Bourret, G. W. Ecoleston,

H. 0. Menlove and L. G. Speir3 Los Alamos Scientific Laboratory and

R. V. Studley

E. I. duPont de Nemours and Co. , Savannah River Plant

Aiken, South Carolina 41

Session II - May 15, 1978, 1:00 p.m.

Chairperson -A. M. Delmastro

Allied Chemical Corporation

Gamma Spectrometric Methods for Measuring Plutonium

R. Gunnink (Invited)

Lawrence Livermore Laboratory, Univ. of Calif 49

'. Rapid Nondestructive Plutonium Isotopic Analysis

J. E. Eager and F. P. Brauer

Battelle, Pacific Northwest Laboratories, Richland, Washington 64

j5. Non-Destructive Assay of Leached Hulls in a Nuclear Fuel Reprocessing Plant

K. J. Eofstetter, B. C. Henderson, J. H. Gray and G. A. Huff (Invited)

Allied-General Nuclear Services, Barnwell, South Carolina 71

+. Safeguards and Nonproliferation

Samuel C. T. McDowell

Department of Energy, Washington, D.C 78

5. Decision Analysis for Dynamic Accounting of Nuclear Material

J. P. Shipley

Los Alamos Scientific Laboratory, Los Alamos, N.M 83

V
6. Analytical Chemistry Measurements Quality Control

Program Using Computer Applications

J. P. Clark and G. A. Huff

Allied-General Nuclear Services, Barnwell, South Carolina ...

Session III - May 16, 1978; 8:30 a.m. Chairperson - W. Pickles

Lawrence Livermore Laboratory

1. Actinide Determination by Rockwell Hanford Operations'

Analytical Laboratories

G. Burch (Invited)

Rockwell Hanford Operations, Richland, Washington

2. X-ray Diffraction Sample Holder for Radioactive

Samples or Samples that React with Air or Light

H. W. Dunn

Oak Ridge National Laboratory, Oak Ridge, Tennessee

3. Feasibility Study of a High-Precision XRF System

for Analysis of Solutions and Slurries

C. R. Hudgens and B. D. Craft

Mound Facility, Miamisburg, Ohio

4. Establishing the Traceability of a Uranyl Nitrate

Solution to a Standard Reference Material

J. P. Clark and C. H. Jackson

Allied-General Nuclear Services, Barnwell, South Carolina ...

5. Preparation of Uranium Standard Solutions for

X-Ray Fluorescence Analysis

C. M. Wong, J. L. Cate and W. L. Pickles

Lawrence Livermore Laboratory, Livermore, California

6. Use of a Non-Linear Method for Including the Mass Uncertainty

of Gravimetric Standards and System Measurement Errors in the

Fitting of Calibration Curves for XRFA with Freeze-Dried

UN0 3 Standards

W. L. Pickles (Invited). J. W. McClure and R. H. Howell

Lawrence Livermore Laboratory, Livermore, California

Session IV - May 16, 1978; 1:00 p.m.

Chairperson - H. C. Whitehead

New Brunswick Laboratory (DOE)

1. Future Approaches to Material Control and Accounting

T. S. Sherr (Invited), G. D. Smith and L. F. Wirfs

Division of Safeguards , Nuclear Regulatory Commission

2. Evaluation of Analytical Capabilities for Accurate

Prediction of Isotopic Correlation Ratios

E. H. Ottewitte

EG&G Idaho, Inc., Idaho Falls, Idaho

3. In-Situ Transuranium Element Measurement Technique

for Wastes Associated with Power Reactor Fuels

K. K. Nielson, R. L. Brodzinski and N. A. Wogman

Battelle, Pacific Northwest Laboratory, Richland, Washington

4. Experiences with a Counter for Plutonium in Crates

R. A. Harlan

Rockwell International, Golden, Colorado

vi
May 17, 1978; 8:30 a.m. Chairperson -A. J. Busah

New Brunswick Laboratory (DOE)

Requirements for Near-Real-Time Accounting of Strategic PAGE

Nuclear Materials in Nuclear Fuel Reprocessing

E. A. Eakkila, D. D. Cobb, R. J. Dietz, J. P. Shipley and D. B. Smith (Invited)

Los Alamos Scientific Laboratory, Los Alamos, New Mexico 208

Verification of 235U Enrichment in SNM Receipts at the

DOE Feed Materials Production Center

H. W. Humphrey

National Lead Company of Ohio 221

Accurate Determination of 5-25 mg of Uranium by Redox Titrimetry

S. D. Reeder and J. R. Delmastro

Allied Chemical Corporation, Idaho Falls, Idaho 247

In Situ Quantitative Determination of Transuranic Elements

in Areas of High-Level Gamma Radiation

R. L. Brodzinski and N. A. Wogman

Battelle, Pacific Northwest Laboratory, Richland, Washington 253

Neutron Correlation Counting for the Nondestructive

Analysis of Nuclear Materials

M. S. Zuokev (Invited)

Brookhaven National Laboratory, Upton, Long Island, New York 261

An Instrument for Monitoring the Transuranic Content of

Chopped Leached Hulls from Spent Nuclear Fuel Elements

N. A. Wogman, R. L. Brodzinski and D. P. Brown

Battelle- Northwest Laboratories, Richland, Washington 284

vii
1
The Chemical and Isotopic Analysis of Uranium, Plutonium and

Thorium in Nuclear Fuel Materials

C. E. PIETRI, J. S . PALLER and C. D. BINGHAM

ew Brunswick Laboratory, U. S. Department of Energy, Argonne, Illinois

ABSTRACT

The New Brunswick Laboratory (NBL) has developed or modified and used highly precise and accurate methods for

the determination of uranium, plutonium and thorium content

and isotopic abundances in a wide variety of nuclear fuel

cycle materials in support of nuclear safeguards programs.

The dissolution, separation and subsequent precise chemical

analysis of these materials with an accuracy of 0.1% or less is achieved using gravimetric, titrimetric,

s pec trophotometr i c and coulometric techniques. Isotopic

abundance measurements on uranium and plutonium are accurately measured with a relative limit of error (95%

confidence limit) of approximately 20 to 0.007% for the

isotope weight range of 0.001 to 97 weight percent using

thermal ionization mass spectrometers. Other methods such

as isotope dilution mass spectrometry, fluorimetry and

radiochemical analysis are used where the uranium and

plutonium sample content is in the microgram range. Many

of the analytical methods used at NBL are in various stages

of automation or mechanization to provide greater

efficiency and productivity at existing levels of accuracy

and precision. Quality assurance programs employing

unknown control standards for the analysis of nuclear

materials are used at NBL to maintain a high level of

reliability

KEYWORDS: Chemical analysis; isotopic analysis; plutonium

assay; uranium assay; thorium assay; dissolution;

chemical separations; automation; quality

assurance programs

INTRODUCTION

During its history, the New Brunswick Laboratory (NBL) has

developed or modified and used highly precise and accurate methods

for the determination of uranium, plutonium and thorium content

and isotopic abundances in a wide variety of nuclear fuel cycle

materials. Currently, these capabilities have been used to

provide comprehensive analytical chemistry support for the

Department of Energy (DOE) and the Nuclear Regulatory Commission

(NRC) nuclear safeguards programs.

NBL in its role as the USA Safeguards Laboratory receives

1
nuclear materials for analysis not only from domestic sources (DOE

contractors and NRC licensees) but from the international

community via the International Atomic Energy Agency (IAEA). The

laboratory receives a complete spectrum of sample types ranging

from product quality uranium, plutonium and thorium metals, their

alloys (binary, ternary or more complex), oxides (prepared over a range of temperatures, or unsintered), carbides, nitrides and

others, to complex scrap materials. This latter category,

consisting of fuel element rejects, process residues, or wastes,

contain a broad range of matrices and have usually undergone

treatments which make them difficult to dissolve under ordinary

conditions. The challenge to the New Brunswick Laboratory is to

accurately measure the uranium, plutonium or thorium content and

isotopic abundances in these widely diverse materials and to

complete these measurements in a timely manner.

The dissolution, separation and subsequent precise chemical

analysis of these materials with an accuracy of 0.1% or less, is

achieved using gravimetric, titrimetric, spectrophotometric and

coulometric techniques. Isotopic abundance measurements on uranium and plutonium are accurately measured with a relative

limit of error (95% confidence limit) of approximately 20 to

0.007? for the isotope weight range of 0.001 to 97 weight percent

using thermal ionization mass spectrometers. Other methods such

as isotope dilution mass spectrometry, fluorimetry and

radiochemical analysis are used where the uranium and plutonium

sample content is in the microgram range.

Because rapid turnaround times are required for the

analytical results from these analyses, NBL has employed automated

or mechanized equipment to accelerate productivity without

sacrificing reliability.

SAMPLE DISSOLUTION

The variety of sample types received at NBL for analysis

require flexible dissolution schemes such as those summarized in

Tables I, II and III. For many years, these techniques have been

the most useful in preparing samples suitable for subsequent

measurement. No one method has been found to be successful under

all circumstances especially when many sample types, regardless of

their generic classification, contain unknown constituents which

complicate or even defeat their proposed dissolution treatment.

In many instances, the analyst resorts to a combination of these

techniques. Currently, lithium-based fluxes for the fusion of

uranium samples 1 and an adaptation of the Los Alamos Scientific

Laboratory low-pressure dissolution technique for plutonium

materials are being evaluated. The general philosophy at NBL is

to completely dissolve all solid samples to produce a\ clear,

residue-free solution. Sample solutions received with sediment

require a separation and dissolution of this residue followed by

recombination to produce a homogeneous solution prior to sampling.

The same logic applies to any multiphase sample. Any organic

material in a sample must be destroyed and the sample

reconstituted prior to analysis. Solid waste or ash samples must

be homogenized and split prior to analysis or the entire sample

2
must be dissolved. It is also required that, for highest

accuracy, all sample solutions be made up by weight and that

appropriate weighed aiiquants be taken for analysis. Proper

sample dissolution is the basis for the reliable measurement of

the element of interest in the material.

SAMPLE ASSAY

Since samples received at NBL are usually of complex or unknown composition, the measurement method must be insensitive to

interferences or must employ techniques to remove interfering

impurities. NBL uses both approaches: (1) a modification of the

Davies and Gray titrimetric method which is highly specific for

the determination of uranium in the presence of most impurities,

(2) a subsequent modification of this method using a tributyl

phophate extraction/titration system for uranium which eliminates

the few remaining interferences and (3) a system incorporating an

ion exchange separation of impurities followed by the controlled

potential coulometric determination of plutonium. Methods for

thorium usually require a separation using precipitation or solvent extraction techniques prior to the determination of

thor i um

.

a. Determination of Uranium

The method originally proposed by Davies and Gray is based on

the titration of uranium (IV) with potassium dichromate in

phosphoric acid media. 3 The NBL-modified method ' ^ is as follows:

A HN0o> HC10|| or H^SOn solution containing up to 150mg U in a 15-

ml volume, or less, is treated with the following reagents and

swirled after each addition: 5ml of I . 5M sulfamic acid, 40ml of

H^POjj (containing 2 drops of 2% KpCr 2 0

7 ), 5ml of 1M_ FeSO^ (10$

H

2 S0 4 solution), 10ml of oxidant (8M HNO3/O.IM sulfamic acid/0. 4$

ammonium molybdate) after 30 seconds wait, 100ml of water after 3 minutes wait, and 125mg of solid vanadyl sulfate dihydrate. A magnetic stirring bar, a platinum wire indicator electrode and a calomel reference electrode is inserted into the beaker, and the

solution, with stirring, is titrated immediately with 0.03N_

potassium dichromate to a potential of 590mv. The amount of

uranium titrated is measured by the quantity of dichromate used to

reach the end point.

The potential interferences associated with this method have

been examined in great detail. Every element in the periodic

table, with certain obvious exceptions, has been experimentally

tested as a potential interferent. An element is considered an interferent if its presence in a concentration of 15$ or less of

the uranium present introduces an analysis error greater than 0.1$

relative. Figure 1 summarizes the results of these studies at

NBL. Note that few elements interfere and most of the

interferences are removed by sample treatment. The remaining

interfering elements can be removed by a separation described

below. High alpha activity equivalent to 51$ ^^^U does not

interfere. The sample treatments which remove interferences are

3
summarized in Table IV.

During the course of these interference studies, the

existence of a nearly universal method for eliminating interfering

elements from samples was developed. The method involves

converting the sample to a solution 4-5 N_ in nitric acid and

extracting the uranium with 30% tributyl phosphate in carbon

tetrachloride. The presence of diverse ions, chloride, sulfate,

perchlorate in reasonable amounts does not adversely affect the

extraction. Large amounts of AKNO^)^, NaNO-, or NaCl also can be

tolerated. The only interfering elements no£ effectively removed

by the extraction are iodine, technetium and gold. The CCl^-TBP

phase from the extraction may subsequently be titrated like an aqueous sample with the same high precision and accuracy.

Under routine conditions, uranium can be measured without

bias using the NBL titrimetric method to a precision of 0.08%

relative standard deviation (RSD) for > 50mg U samples, to 0.15?

RSD for 15-50mg U samples and to 0.25/6 RSD for samples containing

< 15mg U

As indicated previously, the need to increase productivity so as to report analytical results in a timely manner prompted the

mechanization and automation of the uranium titration method.

Four generations of autot itrator s have evolved at NBL. The last

two are currently in operation at the laboratory. One system

simply automates the manual titration although the standard

potassium dichromate titrant addition is measured gravimetrically^

instead of vo lumetr i cally . Studies have shown that this system

can provide unbiased results with a precision of < 0.1% RSD, can process 36 determinations in an 8-hour day and is unaffected by room temperature changes of several degrees. The daily sample

output is limited by the turntable which has a capacity for only

12 samples and must be reloaded manually. The second operating

system, however, is based on the internal electrogeneration of

vanadate ion, V(V), which in essence is the titrant , and can accommodate up to 44 samples and standards in a single loading.

The system is controlled by a minicomputer with an internal

quality control program and fault condition program to provide

high reliability by preventing incorrect measurements. This

autotitrator , developed in collaboration with the Lawrence

Livermore Laboratory', can on a continuous basis compute and print

the uranium content of each sample, including control standard,

which it has measured. Current studies indicate that in the 40mg

U sample range, accurate results can be achieved with an RSD of

0.1%.

b. Determination of Plutonium

To date, no measurement system for plutonium has been

developed which has the specificity that the NBL titrimetric

method has for uranium. The presence of iron, chromium,

manganese, vanadium, Pu(VI) or complexing ions in the generally

accepted potentiometri c , amperometric or controll ed-potential

coulometric methods for plutonium requires corrections,

separations or adjustments for these potential interferences. In
order to eliminate known as well as unsuspected interferences

(including organic material and salts from sample fusion

treatments) NBL policy is to purify plutonium in all samples

received using an ion exchange procedure prior to assay by

controlled-potential coulometry. The basic procedure for this

system is as follows 1 ^: An aliquot containing about lOmg Pu in LN_ HNO^ is reduced to the Pu(III) oxidation state with excess Fe(II)

and the solution is then adjusted to 8N_ HNO^ which produces Pu(IV)

and oxidizes the excess Fe(II). The sample is passed through an anion exchange column (Dowex-1, X-2, nitrate-form resin),

impurities are washed through with 8N_ HNOo and the purified

plutonium is eluted with 0.35N_ HC1-0.01N_ HF solution. After

fuming with HpSO^, the plutonium is determined by an improved

rapid controlled-potential coulometric method^ using a working

gold electrode in 1N_ HN0n-0.02M NH 2 SOoH media. The plutonium is

reduced to Pu(III) and then oxidized to Pu(IV), blank corrections

are applied, and the total quantity of plutonium is calculated

using an electrical calibration factor.

Routinely, 6-12mg Pu samples can be assayed without bias with

a precision of 0.06? RSD using this system. Additional data using

NBS SRM 9^9e plutonium metal and SRM 9^4 plutonium sulfate

tetrahydrate standards in 135 determinations over a period of one year have verified this performance, Table V.

The need for automating plutonium analysis concurrently with

uranium analysis has been recognized and the mechanization of the

NBL plutonium analysis system has been underway for several years.

An apparatus (Autosep) 1 ^ has been developed in which groups of ten

plutonium samples are automatically separated from impurities

according to the manual ion exchange method previously described.

The purified plutonium is subsequently processed through an apparatus ( Autocoulometer

)

1 3 designed to automatically and

continuously analyze two groups of 50 samples (total 100 samples)

by the NBL controlled-potential coulometric method. A 5-fold

increase in sample throughput is anticipated with this automated

system while retaining an accuracy and precision of < 0.1%.

c. Determination of Thorium

NBL uses the classic gravimetric technique whereby an excess

of saturated oxalic acid is added to a boiling solution containing

at least 200mg Th. The solution is allowed to digest overnight,

the precipitate is filtered and then ignited at 1100°C to constant

weight. A precision of < 0.1$ is obtained; results can tend to be

slightly biased probably due to the precipitation of certain

oxalates if metallic ions are present in large amounts. In some instances a prior precipitation of the hydroxide with fresh NH^OH

is necessary to remove interferences such as fusion salts. An

alternative titrimetric method using ethylenediaraine tetraacetic

acid (EDTA), has been used in which a solution buffered at pH 2.5

and containing about iOOmg Th is titrated with 0.01M EDTA solution

(standardized against copper) to a xylenol orange indicator end

point. A comparison with the gravimetric method gave an RSD of

0.08$ without significant bias. A third method utilizing

spectrophotometry consists of mixing 0.1$ Thorin solution (2- [2-

5
hydroxy- 3 , 6-disulf o-l-naphthylazcj ben zenearsoni c acid) with a solution containing 0.5 to 2mg Th. After dilution and pH

adjustment to 0.8 +_ 0.1, the absorbance is read at S^nm.

Unbiased results with a precision of 5-10% are usually obtained.

However, since so many interfering elements usually are present in

the sample, a preliminary separation of thorium is required.

Other Methods

Samples containing small amounts of uranium at the 0.05 to

2.5mg level can best be determined by f luorimetry

1 ^ The

procedure consists of the following: Sample solutions are aliquoted into centrifuge tubes, the I^O^ group is precipitated

using a solution of saturated ammonia gas, the supernantant is

decanted, the precipitate is dissolved with nitric acid and the

^2^3 S roa P i- s reprecipitated . The precipitate is dissolved with

nitric acid, magnesium nitrate hexahydrate is added and dissolved.

The uranium is extracted with ethyl acetate, one milliliter of the

extact is pipetted onto a gold dish, the dish is ignited and fused

with a sodium f 1 uor i de/ sod ium-potass i um carbonate flux. The

fluorescence is measured with a fluorimeter and the uranium

content is determined from a calibration curve prepared from

standards processed in exactly the same manner as for samples.

Unbiased results with a precision of < 10% RSD have been obtained.

Another method for the analysis of small amounts of sample,

namely, the isotope dilution mass s pectrometri c technique, has

been used at NBL for the determination of microgram amounts of

Plutonium and uranium in d isso 1 ver- type solutions^ in which the

uranium to plutonium ratio is typically 200 to 1. In this

procedure, samples containing about 10 /<g Pu and 2mg U are spiked

with known amounts of ^ Pu and ^33[j tracers, equilibrated

physically and chemically, and the uranium and plutonium fractions

are separated by a version of the anion exchange method described

for plutonium assay. Each fraction is analyzed by mass spectrometry and the plutonium and uranium content in the sample

is calculated. Experience at NBL has shown this method to be

unbiased with a precision of about 0.5% RSD, although data under

optimum conditions have been obtained with a precision of 0.2%.

This method is presently being evaluated in long-term studies to more accurately define its reliability.

ISOTOPIC ABUNDANCE MEASUREMENTS

The isotopic composition of uranium and plutonium samples is

routinely determined using thermal ionization mass spectrometry.

In general, only high purity uranium and plutonium can be reliably

analyzed with the thermal ionization technique. Impurities,

especially alkali elements, tend to produce unstable emission and

may alter the observed ratios in an unpredictable manner. Furthermore, in the case of plutonium analysis, the sample must be

free of ^38^ and 21n Am tQ avo i<j interference in the 2 ^ Pu and

Pu measurements. Hence, in the majority of cases, the uranium

and plutonium must be separated by an appropriate procedure.

6
a. Uranium

Uranium is separated from impurities for measurement by mass spectrometry by precipitation as the hydroxide with NH^OH followed

by dissolution in HNOo and subsequent extraction of the uranium

into diethyl ether. Where thorium is present, the hydroxide

precipitate is dissolved in 8 M HC1 and passed through an anion

exchange column (Dowex-1, X8 , chloride-form resin). The column is

washed with 8 M_ HC1 to remove thorium, and the uranium is eluted

with 0 . 1N_ HNO^. The separated uranium is further purified by the

ether extraction method described above.

The mass spectrometry method is applicable to the

determination of the isotopic composition of uranium with sample

sizes in the range of about 20-30 /"g U. 1 ° Analyses are made with

either a 12" radius 60° sector or a 12" radius 90° sector mass spectrometer equipped with a triple-filament thermal ionization

source, Faraday cup ion collector, programmable calculator data

acquisition system and an NBL-designed vacuum lock. Isotope

ratios as low as 0.02 are measured routinely with a precision of

0.2? at the 95% confidence level. The basic method consists of

loading the purified uranium onto a disposable triple-filament

holder ("bead" or "hat") and inserting it into the ion source assembly of the mass spectrometer. The isotope ratio technique is

used to determine the relative abundances of the uranium isotopes.

The ^35y pea^ j_ s use d as the reference peak and the U, 236y and

^3 u peaks are measured relative to it. Magnetic peak switching

under calculator control is used to bring the 2 35jj p ea k ancj the

isotope being measured alternately onto the collector slit. The

ion currents are amplified with a vibrating-reed electrometer and

measured with a digital voltmeter and variable gain amplifier

interfaced to the calculator. The individual isotope ratios,

relative standard deviations, and the weight percent abundances

are computed and printed out by the calculator. Since the 2 35{j

isotope is usually the major isotope, sufficient peak sets to

yield 15 238/235 ratios and three 23^/235 and 236/235 ratios each

are taken. Since thermal ionization mass spectrometry is not an absolute measurement technique, all sample data are corrected from

the analysis of NBS SRM U-005 through U-970 isotopic standards

under the same controlled conditions as for samples. About 10

samples (including standards) can be run daily using the vacuum

lock to reduce sample change time. Typical precision over the

entire isotopic weight percent range is shown in Table VI.

b. Plutonium

Plutonium is separated from impurities for mass spectrometric

measurement by anion exchange using a miniaturized version of the

technique previously described for plutonium assay. Where large

amounts of uranium are present in the original sample (PUO2UO2

material, for example) the ^ Pu content is often determined by

alpha spectrometry on a separate fraction, or a two-stage ion

exchange separation is performed. (Even traces of 2 ^ U in the Pu

fraction give apparent high 2 ^ 8 Pu values especially when the

7
latter isotope is only abundant to <0.1%). Occasionally, traces

of neptunium as 2 3?Np may be found in the plutonium fraction since

neptunium has similar characteristics in this separation

procedure. The mass spectrometry method is applicable to the

determination of the isotopic composition of plutonium with sample

sizes in the range of about 0.5 to 5 /j g Pu. 1 ^ These relatively

large sample sizes are used not because of instrumental

limitations, but rather to minimize environmental and reagent

contamination in the sample preparation. Analyses are made with a 12" radius 90° sector tandem mass spectrometer equipped similar to

the uranium instruments but utilizing an integrating ratiometer in

the data acquisition system. As with uranium, isotope ratios down

to 0.02 can be measured with a precision of 0.2% at the 95%

confidence level. The basic operating method is similar to the

uranium procedure. In determining the relative abundances of the

plutonium isotopes, the 2 39p u peak i S used as the reference peak

and the 2 ^ Pu, Pu, 2 Pu and 2 Pu peaks are measured relative

to it. Accelerating voltage switching under calculator control is

used to bring the 2 ^^Pu peak and the isotope being measured

alternately onto the collector cup. The ion currents are amplified with a vibrating-reed electrometer and measured with an integrating ratiometer interfaced to the calculator. Data

computations are performed similar to the uranium manipulations.

Since the normal isotopic distribution , in decreasing abundance,

is 239 Pu, 240 Pu, 241 Pu, 242 Pu and 23 °Pu, sufficient peak sets to

yield 22 240/239 ratios, 18 241/239 ratios, 14 242/239 and 238/239

ratios each are taken. Plutonium data are corrected using NBS

uranium isotopic standards and the validity of these corrections

are verified using the NBS SRM 946, 947 and 948 plutonium isotopic

standards. About six samples and standards can be run daily.

Typical precision over the plutonium isotopic weight range is

shown in Table VII.

QUALITY ASSURANCE PROGRAMS

In order to assure reliable measurements, analytical

laboratories incorporate measurement control systems into their

sample analysis schemes. The NBL Quality Assurance Program is

based on the use of unknown standards to monitor the performance

of the analyst. The analyst uses the procedure shown in Figure 2 to implement the program. He enters the value found for the

measurement of an unknown standard into a computer terminal and

receives an evaluation of his performance as "in control" or "out

of control." ("Out of control" responses require remedial action

to identify and resolve the discrepancy). In order to maintain

the integrity of the system, no corrections or changes may be made

once a result is entered into the computer. The computer is also

programmed to automatically update the analyst's control limits.

Current control limits (90% confidence level) for uranium are:

< 15mg U, 0.5%; 15-50mg U, 0.30%; > 50mg U, C.20%; and for

plutonium are: 6-12mg Pu, 0.18%. All analysts must be certified

before routinely analyzing samples by demonstrating their

proficiency within the accepted control limits.

8
SAMPLE ANALYSIS AND THE REAL WORLD

Most of the methods previously described have been used

successfully at NBL for many years. All methods whether new or old have been thoroughly tested and certified to be highly

reliable on a great variety of sample material. However,

regardless of the proven superiority of these methods of analysis,

NBL, as well as most analytical laboratories, has been faced with

the enigma of applying highly accurate techniques to inadequately

sampled material. Shipper-receiver differences may result, in

many instances, not so much from differences in laboratory

capability, techniques, or methods, but from the failure of both

parties to analyze exactly the same sample. Many samples received

at NBL are grossly heterogeneous and it is quite evident that the

"average" sample is far from representative. The work of

homogenizing this type of sample after receipt by calcining,

blending or reconstituting, or alternatively, by using the entire

sample to avoid processing, can be a lengthy and tedious

procedure. Moreover, the subsequent precise analysis of the

processed sample gives results that have little meaning when

related to the larger material source from which it was taken.

Not only must adequate sampling and sampling plans be addressed

but necessary attention should be paid to the proper packaging and

handling of samples so that both the shipper and receiver are analyzing the same material. For example, the problem of shipperreceiver differences for hygroscopic low-fired PuO^ has existed

for more than 10 years and only now are efforts being made to

standardize fundamental sampling, packaging and handling

conditions

The New Brunswick Laboratory perceives as its mission not

only the performance of high quality chemical and isotopic

analysis of nuclear materials but also the identification and

resolution of analysis-related problems in order to provide

meaningful support to the USA Safeguards Program.

REFERENCES

1. Unpublished data, New Brunswick Laboratory.

2. J. W. Dahlby, R. R. Geoffrion and G. R. Waterbury, USAEC

Report LA-5776 (1975).

3. W. Davies and W. Gray, Talanta 11, 1203 (1964).

4. A. R. Eberle, M. W. Lerner, C. G. Goldbeck and C. J. Rodden,

USAEC Report NBL-252 (1970).

5. USAEC Report NBL-272, Appendix A (1974).

6. L. Z. Bodnar, M. W. Lerner and J. M. Scarborough, USAEC

Report NBL-272, 12 (1974).

7. B. W. Moran, USDOE Report NBL-288, in preparation.

9
8. C. G. Goldbeck and M. W. Lerner, USAEC Report NBL-265,

5 (1972).

9. J. E. Harrar , W. G. Boyle, J. D. Breshears, C. L. Pomernacki,

H. R. Brand, A. M. Kray, R. J. Sherry and J. A. Pastrone,

USDOE Report UCRL-52351 (1977).

10. J. R. Weiss, C. E. Pietri, A. W. Wenzel and L. C. Nelson, Jr.,

USAEC Report NBL-265, 94 ( 1972).

11. M. K. Holland, J. R. Weiss and C. E. Pietri, Anal. Chem. 50,

236 (1978).

12. J. R. Weiss, A. W. Wenzel and C. E. Pietri, USAEC Report

NBL-272, 74 (1974).

13- G. E. Peoples, S. A. Malone, J. R. Weiss, A. W. Wenzel and

C. E. Pietri, USAEC Report NBL-277, 80 (1976).

14. C. J. Rodden, Analysis of Essential Nuclear Reactor

Materials (Superintendent of Documents, Washington, D. C. t 1964), Chapter 1, p. 132.

15. C. E. Pietri, A. W. Wenzel, L. J. Jasper, Jr., H. 0. Finley

and L. C. Nelson, Jr., USAEC Report NBL-238 (1967).

16. E. L. Callis, New Brunswick Laboratory Report, in

preparation

17. R. L. Hague, New Brunswick Laboratory Report, in

preparation

18. C. E. Pietri, E. H. Jacob and J. T. Bracey, USAEC Report

NBL-276 (1975).

10
TABLE I. DISSOLUTION OF URANIUM MATERIALS

Materials

00- u

3 o

8 UF:

Dp powders and

ellets, ammonium

iuranate, wastes,

D 2 -Th0 2

, etc.

re concentrates

TGR fuel beads:

SiC-coated, pyrolytic

ar bon-coated

Th carbide)

C, uc 2

Treatment

Dissolve in HNO ^

a

Dissolve in HNO,, filter, fuse

residue in NaHSO^ or Na^O^.

combine solutions. 3

Dissolve in HNOo, fume sample

aliquants with F^SO^ and HF

1. Ignite to remove carbon, fuse

with Na 2 C0o, dissolve cake and

fume with H^SO^ and HF to remove

silica. Precipitate U with

NH^OH to remove excess fusion

salts, dissolve residue in HNO^. 3

2. Ignite to remove carbon, treat with

Cl

2 at 900°C to decompose SiC,

ignite as in 1, dissolve as with

U0 2 -ThO2

• Ignite, dissolve in HNO^, fuse

any residue. 3

Al, U-Si, UOp-SS Dissolve in HCl-HNO^. fume with

HCIO^, filter. Volatilize

silica with HF , fuse remaining residue

with Na-,C0 o . b

'issium alloy

'issium dross

sh samples

Dissolve in HC1-HN0- treat residue

J0

2 -Zr0

2 -Nb-Zr , J0

2 -BeO

Dissolver solutions,

organic solutions,

wastes

U-Zr

with NaOCl and NaOH , acidify with

HC1, combine solutions. 15 (Residue from

dross requires fusion with NaOH).

Fuse with Na 2 C0o-NaN0o (5:1) and/or

NaHSOjp dissolve cake in HNO^-HClOi,,

volatilize silica with HF , fume with

HC10 4. a

Dissolve in HN0o-HF. a,b (Large

quantities of N5 may be removed by

precipitation with S0 2

).

Homogenize and reconstitute multiphase

mixtures where necessary. Destroy

organics with hot H^O^-HNO^

a

Dissolve in HF. a ' b

NOTES: a Fume sample aliquants with H^O^

b Fume sample aliquants with HCIO^

11
TABLE II. DISSOLUTION OF PLUTONIUM MATERIALS

Material Treatment

Pu, Pu-Al, Pu-Ga

U-Pu-Mo

Pu0 2 , Pu0 2 U0 2

Pu0 2

, fired > 800°C

(U-Pu)C

6 N_ HC1 or 18 N_ H

2 S0 i|

3 N HC1 - 8 N HNOj - 0.1 N HF

8 N_ HNOo - 0.1 N HF, or fuse

in NaHSOjj

Fuse in NaHSO^, or 8 N HNOg - 0.1 N_ HF (extended heating;

Ignite; 8 N HNO ^ - 0.1 N_ HF , or fuse in NaHSO,,

"Calcined ash"

"Brick residues"

(Al 2 0o, MgO , CaO

Fe 2 0,, Si0 2

)

Leach in 8 N HNOo - 0.1 N_ HF , fuse residue in NaHSOjj, or fuse in

NaHSO^

Fuse in NaHSOjp dissolve in

2-8 N_ HNOo, fuse residue in

Na 2 C0

3 "Grinder sludge" : (SiC) Fuse in NaHSO.

TABLE III DISSOLUTION OF THORIUM MATERIALS

Material Treatment

Metal 8-16M HN0

3 + k% HF , or

6N HC1 (F present).

Silica-base ores (thorite, monazite)

NaHF 2 fusion, or hot

concentrated H-pSO^ (losses occur ) .

Th0 2

, low-fired

( < 600°C)

S-15M HN0

3 (F~ present)

Th0 2

, high-fired

( > 600°C)

16M HN0

3 + 0.03M HF with

extended digestion (< 3 days)

or NaHSO^ or Na 2 S 2 0

z fusion, or HCIO^ (F present)

f umi ng

U0 2 -Th0 2 See Table 1

12
TABLE IV. SAMPLE TREATMENTS WHICH REMOVE INTERFERENCES

TO THE NBL TITRIMETRIC MEASUREMENT OF URANIUM

Element Treatment

Ao , Tc, Ru, Os, Fume sample aliquant in 3-5ml HCIO^

31, Br, large excesses and/or H2SO^

of F

I Add bromine water, evaporate, fume in

HC10 4 or H

2 SOjj

Is, Sb, Sn Add small amount of F^C^Oy to

sample aliquant prior to Fe(II)

reduction step

Mn, Mo, Ru, Pd, TBP Extraction in CCl^ 3

\g, Os, Ir, Pt, Hg,

\s, Sn, Sb , CI , Br

L0$ Au Reduce to metal and separate

Jote: All of the interfering elements are listed in

this table.

Kerosene interferes causing results low by 20-40% and cannot

be used as a solvent for TBP in this application. Cyclohexane

atlsfactory substitute for CC 1 ^

13
TABLE V. DETERMINATION OF PLUTONIUM USING RAPID CONTROLLED

POTENTIAL COULOMETRY

P reparation

Standard

Type

No. of

aliquots a Mean Recovery,

% RSD,%

A 949e 5 99 .96 0 . 07

B 949e 11 100. 03 0 . 06

C 949e 11 99 .99 0 . 06

D 949e 9 99 .99 0.03

E 944 5 100.02 0.05

F 944 17 100 .01 0. 07

G 944 39 100 . 01 0. 06

H 944 11 99 .98 0.05

I 944 7 100.03 0 . 04

J 944 18 100. 02 0. 05

K 944 2 100 . 00 0. 01

Total 135 100. 01 0. 06

a 6-12mg Pu aliquots.

14
TABLE VI. RELATIVE LIMIT OF ERROR FOR URANIUM ISOTOPIC ANALYSIS

(95$ Confidence Limits)

LiT

• PC* •

f *

n ^ _ o 7 nu u n 7 _ i n U . 1 — J. . u

AU

O

J

L . U — J. . D

nu

X . — -1- J un c U

1 n.i c

X J — c. \J 0 1 6 ?£1 nU _— ?£ ^ 0 ^} ju 1 0 1 ?

*r u — n J

n<j x 1 n 1 0 0 OQ\j y 0 08

n _ f\ r n 07

^ R _ Q 0 \J U Q - 7 ? 1 J 0 7 3 - 7Q 0 04

79-85 0 03

85-88 0 025

88-90 0 020

90-91.

0 015

91.5-93 0 013

93-94 0 011

94-95 0 010

95-96 0 009

96-97 0 008

> 97 0 007

15
TABLE VII. RELATIVE LIMITS OF ERROR FOR PLUTONIUM ISOTOPIC

ANALYSIS (95% Confidence Limits)

Wt . % Pu

0. 001-0. 01

0.01-0.03

0.03-0.05

0 . 05-0 .

0 . 1-0 . 5

0 . 5-1

1-5

5-10

10-15

15-20

20-25

65-70

70-75

75-80

80-85

85-90

90-95

> 95

L.E. , %

20 . 0 10.0

5 . 0

2 . 0

1 . 0 0.5

0 . 2 0.18

0.16

0.14

0.12

0 . OH

0 . 035

0 . 03

0 . 025

0 . 02

0.015

0.01

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18
Analytical Methods and Laboratories for Safeguards

and Accountability: Present and Future

by

L. T. CORBIN

Oak Ridge National Laboratory, Oak Ridge, Tennessee 37830

ABSTRACT

Accountability for special nuclear materials at ORNL requires a number

of different analyses depending upon the kind and source of the material.

The determination of uranium and plutonium in highly radioactive reactor

fuel dissolver solutions must be performed remotely in hot cells or in

glove boxes after a separation of the heavy elements from the bulk of the

fission products. These determinations can be made by several different

chemical techniques or radioactive solutions may be measured simultaneously and

accurately by isotope dilution mass spectrometry.

Measurements where accuracy is essential are subject to a quality

assurance program for all phases of the operation: equipment, standards,

sampling, aliquoting, documentation and reporting.

Special material samples are stored either in hot cells or in glove

boxes for analyses. The analytical laboratories are in zones that are not readily accessible to unauthorized personnel.

Future safeguards analytical laboratories will operate under the

premise that if samples always remain in the hot-cell and glove-box

systems, diversion of material will be difficult. Analytical procedures

will be highly automated to minimize the need for human sample handling,

and sample storage and retrieval will be under computer control.

Hot cells and glove boxes must and will be designed to optimize

automated operations. Thus, they will have a very different appearance

from hot cells of today.

Analytical techniques not previously used for automated radioactive

operations must be adopted. Spark-source mass spectrometers are already

being converted for hot cell use. Thus, there will be many opportunities

for new and innovative approaches to chemical and radiochemical techniques in the years to come.

KEYWORDS: Accountability; analyses; remote; plutonium; safeguards; uranium

19
An Automated X-ray Spectrometer for Mixed Oxide Pellets

by

M.C. LAMBERT, M.W. GOHEEN, M.W. URIE, and N. WYNHOFF

Hanford Engineering Development Laboratory, Richland, Washington

ABSTRACT

This paper discusses the development of an energy dispersive x-ray (EDX)

spectrometer for the rapid, automated, close-coupled analysis of solid mixed

plutonium-uranium oxide fuel pellets. Reasons are given for the system design, which is intended to give a relative precision (RSD) of 1 0.2% in a total analysis time of two minutes. The principal problems in an EDX system

are in maximizing the plutonium count rates.

KEYWORDS: X-ray spectrometer, energy dispersion, nuclear fuel assay, in-line,

automation, faster counting.

INTRODUCTION

The Hanford Engineering Development Laboratory has a major responsibility for the deve

opment of fuels for the breeder reactor program. The cost of fuel fabrication represents

major portion of the total fuel-cycle costs for power reactors. Therefore, the earlier i the fabrication process that analytical measurements can be obtained the greater will be

their effect on the quality of the product. To verify the quality of the finished oxide

fuel pellets, chemical analyses are currently performed days or weeks after an entire

fabrication run has been completed. This is too late to affect control of the process

during any current production run. Consequently, we are developing a close-coupled technique for rapid fuel pellet analys

in order to improve fabrication quality control and to decrease or eliminate costly produ

tion reruns. We say close-coupled, rather than in-line, because our present concept is t have the x-ray spectrometer with its own glove box close to, but not in, the production

line. Samples from more than one location in the process can be transported pneumaticall

into the analytical glove box and back again to the production line in an atmosphere of

nitrogen.

The x-ray fluorescence method has been shown to be feasible by a number of people. Som> of them are E.A. Hakkila at Los Alamos \ by Von Baeckmann at Karlsruhe, West Germany2>

W. L. Pickles (the chairman of this meeting) and J. L. Cate, Jr. at Lawrence Livermore

Laboratory^, and by M. C. Lambert at Hanford^, analyzing uranium and plutonium in solutio

1 E. A. Hakkila, R. G. Hurley, and G. R. Waterbury, Appl. Spectrosc. 22, 434-7(1968).

I 0

tut

ft!!

, Ik

it!

fil

2A. von Baeckmann, Anal. Chim. Acta 47, 431-8 (1969)

3W . L. Pickles and J. L. Cate, Jr., Advances In X-ray Analysis (Plenum Press, New York

1974) , Vol. 17, Page 337.

^M. C. Lambert, "X-ray Spectrographic Determination of Potassium, Uranium, and Thorium

in Pyrochemical Samples", HW-65801 (1960)

20
lizlein, Gerding, and Steindler at Argonne National Laboratory^ made a comprehensive

ly of x-ray fluorescence with various forms of the solid mixed oxides. Their work

Lcated great possibilities for the analysis of pressed pellets, but not for bulk mixed

ders. We plan to achieve even better precision and faster analyses of solid fuel pellets.

Our primary goal is to develop an x-ray spectrometric method for the rapid determination

the plutonium and uranium contents of finished fuel pellets immediately after they come

n the sintering furnace. We may also analyze the mixed powder many hours earlier in the

duction process. Powder samples will likely be pressed into a green slug before analysis

Drder to achieve higher and more uniform density. Since the fuel pellets will be fabried from high-exposure plutonium with its increased radioactivity, the spectrometer must

capable of automated remote operation. A secondary goal is to achieve sufficient preion for the analyses to be useful to the Safeguards program by providing an accurate

entory control at one or more locations in the fabrication process.

One of the difficulties in making the analysis both rapid and precise and also in autoing the sample handling is the small size of the pellets. Current fuel pellets are only

• mm long and 4.95 mm in diameter.

The state of the art for commercial x-ray spectrometers in early 1975 was the basis of

;t of our decisions when we drew up the specifications. Since no commercial spectrometer

ild meet our requirements without some design work and modifications we received only ee bids that were responsive. The most practical bid was a combined one from Kevex

-poration and Tracor-Northern, Inc. The bids were received in May 1976, our negotiations

1 discussions about design requirements were completed by late summer, and we began

:eiving portions of the system in May 1977.

Figure 1 shows the nearly complete spectrometer before it was attached to the end of a

Dve box. It is an energy dispersive solid state spectrometer. Kevex did the custom

orication and assembled all the instrumentation. At the left is a Siemens model 805

ray generator, selected in part because it could be automatically controlled by a computer,

day there are other generators that can also be computer controlled. Next to the generar are two cabinets containing 31-liter liquid nitrogen dewars. Two lithium-drifted silicon

tectors and the x-ray source are housed in the metal box above the cryostats. This small

'tal housing was temporarily used in place of a glove box. The inside of this box will

seen in Figure 2. To the right of the detector compartment is a Tracor-Northern TN-11 control and processs g system which includes a 32,000 word memory minicomputer and a dual floppy disc drive.

' e cabinet to the right of that contains a Kevex 5100 microprocessor system. Next is a Cwriter and the last unit on the far right is a Kevex 5100C four-color display and keyard terminal. Kevex had no minicomputer system at the time we purchased the equipment,

we specified a TN-11 control and data processing system from Tracor-Northern. Although

lere were other similar data processing systems available we chose the TN-11 because of

le abundance of good software available from Tracor-Northern, and because of our prior

iowledge of the Flextran language used by the TN-11 plus the fact that we will have to

rite most of the control programs.

Figure 2 shows two detector tubes pointing upward toward the sample position. An x-ray

ibe comes in from the left side. Right above that is a 6-position filter turrent, and

ehind that is a 6-position secondary target turret. The sample pellet can be excited

Lther by the x-ray tube directly or by selectable secondary radiation. Figure 3 shows the

ame housing as seen from the rear, showing it bolted to the end of the glove box. We will

ime back to these pictures later. But now the question is: Why did we design the system

ike this?

CHOICE OF ENERGY VERSUS WAVELENGTH DISPERSION

We had had more than twenty years experience with wavelength dispersive x-ray fluorescence

WDX) , but very little first hand experience with energy dispersive (EDX) equipment at the

ime we were setting the specifications. It seemed to us that the EDX system had sufficient

esolution for this application and offered a great deal in simplicity in that a single

mall detector could simultaneously collect data from two or more peaks and two or more

ackgrounds. If the working end of the instrument were to be put inside the glove box an DX system would take up much less space. Furthermore, since an EDX system utilizes no oniometer it would be much simpler to automate.

5j. G. Schnizlein, T. J. Gerding, and M. J. Steindler, "Application of X-ray Fluorescence

to On-Line Analysis of Plutonium in Mixed-Oxide FBR Fuels", ANL-8004 (1974).

21
Leak

jkr

Ikk

ti

till

III!

'11

he

;i

A wavelength dispersive system would lose the advantage of its higher count rate capa- tj s bility if four or more sequential measurements were to be made with a single goniometer. Oi the other hand four or more separate mechanical goniometers and detector systems, which

would permit simultaneous counting of the plutonium and uranium Ia j lines and backgrounds,

would crowd the glove box, require frequent tuning and calibration of the individual spectrometer systems, require more frequent use of standards, and require much more attention

from the operator.

Consequently, we chose to use the American-created solid-state detector and are trying

to push the energy dispersive technique to its current maximum capability.

CHOICE OF la j VERSUS K^ FLOURESCENCE

We were next faced with the choice of using plutonium and uranium fluorescence at about 14 keV or their K^ ^ fluorescence at 104 keV. Using the Pu I^j line doesn't allow

analysis of as deepa layer in the pellet as we would like to sample. Attenuation calculati

show that only 10% of the Pu I^j fluorescence escapes from a depth of 28 ym from a pellet

consisting of 25% Pu02 and 75% of UO2; 20% of the fluorescence escapes from a depth of 20

and 50% from a depth of only 8.5 um. Thus the effective depth of analysis is less than 20 However, that was thought to be sufficient depth of analysis for two reasons. For one, th

oxide powders are ground to about a 10 micron size; hence there should be no significant

particle size effect. For another, the pellets will come directly from the sintering

furnace in a reasonably inert atmosphere. Even if measurable surface oxidation were to occur it would not alter the plutonium to uranium ratio. Furthermore, use of the low

energy L-lines for analysis greatly simplified design and construction of the glove box,

sample manipulator, and parts of the spectrometer, since only 3.0 mm of stainless steel

provides 12 half-thickness of shielding, giving an attenuation factor of 4000 for 40 KV

x-rays.

Use of the Kai lines would result in an analysis depth about 52 times greater, which

would be good. However, to excite this high energy fluorescence would require an x-ray tu to be operated at about 150 KV. This would mandate the use of a radiography-type generato

instead of the highly stabilized laboratory generator. It would also require some heavy

shielding, e.g., 6.4 mm thickness of lead plus 6.4 mm of stainless steel, to protect perso

from a 150 KV source plus the secondary Pb 1^ fluorescence. Finally, it would not be

possible to utilize secondary radiation, nor even effective filtering. As a consequence

there would be a sloping x-ray background of large magnitude under the analytical peaks

resulting in some loss of precision. Therefore, we decided to use the L-spectral lines of

plutonium and uranium.

CHOICE OF HIGH POWER VERSUS LOW POWER X-RAY TUBE

The recently developed low power (10 to 50 watts), air-cooled x-ray tubes, with selffiltering transmission targets and end window design, are very appealing from the point oft view of space requirements. The extremely stable power supply is only a single chassis,

requiring no extra floor space. Since these tubes have no heated filament wire, they can t be pulsed on and off with extreme rapidity by simply deflecting the internal electron beam: When the amplifier senses the beginning of a count it turns off the x-rays and they remain

off until the ADC is no longer busy processing the signal. This decreases the amount of

signal pile-up significantly and permits a higher count rate. Using one of these pulsed

tubes recently in our general laboratory spectrometer, which has only one detector, we wer able to analyze Pu-U fuel pellets with a relative precision (RSD) of Hi 0.4% in two real

minutes per pellet.

However, we thought 50 watts was not enough power and chose a 3000 watt x-ray tube to permit the use of selectable secondary targets, which produce a minimum of x-ray backgroun

and permit the most efficient excitation of plutonium. The reserve of power also allows

spacing enough between the sample and detector to keep all of the instrument outside the

glove box. This greatly simplifies any maintenance or modification work. The geometry

still turned out to be fairly tight. Distance from the sample to the detector crystal in our present system is only 4.7 cm. Figure 4 shows the sample manipulator pulled back away from the beryllium window inside

the glove box. The pellet rotator is oriented for an end view of a pellet. The housing

has X, Y, Z vernier adjustments and the open side toward the operator has an interlocked

lead-glass window for shielding. The open end of the sample housing fits inside a framewc

shown on the end plate in Figure 4. This prevents x-ray leakage and yet allows the sample

22
-.sing to be moved up or down or sideways. The x-ray source and detector compartment on

! outside, see Fig. 3, bolts to studs on a stainless steel plate, the connection being

ie gas tight by means of a large O-ring.

! The end plate of the glove box contains two 125 um (5 mil) beryllium windows one inch in

imeter. The 5 mm space between the two windows is slightly pressurized and gauged so that

.eak in either window will set off an alarm. The housing for the x-ray source serves as

secondary containment vessel if both windows should ever break. The two windows to-

:'aer absorb less than 2% of the plutonium and uranium fluorescence. The sample manipul-

)r housing inside the glove box and the x-ray tube and detector housing on the outside of

2 glove box are both interlocked to the end plate so that they must be in contact with

a end plate. As a result, x-ray leakage is barely measurable.

PARTICLE SIZE EFFECT AND ORIENTATION OF SOLID PELLETS

: Since 50% of the plutonium and uranium fluorescence originates at depths less than 9

:om the surface of a solid mixed oxide fuel pellet and since the average particle size of

je oxide powders, before pressing and sintering, is about 10 or 11 microns, there is the

issibility for a slight particle size effect.

: We analyzed the cross sectional ends of several fuel pellets on our general laboratory

ectrometer by measuring the Pu/(U+Pu) ratio, then repeating this after rotating the pellet

, 90, and 135 degrees. The maximum relative variation in the ratios was 0.56% whereas

e precision of reproducibility of these ratios under the same conditions was i 0.34%. The

mber of data was insufficient to give statistically positive results but there appeared

be a very slight variation in Pu/U ratios with changes in orientation of the pellets.

Consequently, a sample manipulator was designed so that it could present the pellet

r either an end view or a side view analysis and rotate it at variable speeds. An end

.ew orientation presents an analysis area of 19 mm^. The rotated side view presents the

itire perimeter for about 85% of the length of the pellet and allows analysis of about 80

»2 and this is the way we plan to look at the pellets.

The final version of the sampling device has not been designed yet. It will be completely

ltomated, have two or more standards ready for calibration checks at programmed intervals,

"id pick up samples from an incoming line or boat, then return them after their analysis.

CHOICF OF EXOITTNC X-RAYS

The selection of the excitation source is important for optimizing the analysis of fuel

sllets. One of the chief parameters is the separation between the exciting x-rays and the

[II absorption edge of plutonium. Figure 5 shows the energy values of uranium and plutonium

fluorescence, the absorption curves of uranium and plutonium, and the energies of the

naracteristic radiation from a number of exciting target materials. In fuel containing

5% plutonium dioxide and 75% uranium dioxide the precision for the plutonium measurement

s the controlling precision. Ideally, the exciting radiation must exceed the energy of

he plutonium Ljjj absorption edge, but be as close to it as possible in order to be highly

bsorbed. Thus technetium Ka radiation would produce the greatest plutonium La fluorescence yield. We obtained some technetium metal, since quite a bit of it was separated at

anford, but it is not a practical target material as it emits 2.9 meV beta particles. The

ext heavier metals are ruthenium, rhodium, palladium, and silver. We have been using a

hodiura secondary target for much of our early work. Recently we added a ruthenium target

hich we pressed from 80-mesh pure ruthenium particles, producing a surface in appearance

ike a shiny continuous metal sheet. Since we have no ruthenium foil we use a 0.025 zironium foil to filter out the continuum.

Figure 6 shows the 50 KV exciting spectrum from 0 to 40 keV as scattered by polyethylene,

t includes the Ka and Kg ^ peaks, their incoherently scattered Compton peaks, and the

iremmstrahlung with various degrees of filtering. We filter out much of the continuum.

Spectra from the other targets are similar to that from rhodium with the peaks displaced

ilightly to lower or higher energies.

Figure 7 shows the relative strengths of the rhodium peaks and the Compton peaks when

scattered from matrices of different densities. The huge Compton peak is scattered from

polyethylene. The rhodium and Compton peaks are about the same size when the sample

oatrix is aluminum. The Compton scattering decreases considerably as the matrix is changed

:o heavier and heavier matrices, such as titanium, zirconium, and lead. There is almost

io Compton scattering from uranium and plutonium. The oxygen content of fuel pellets, however, creates a small Compton peak. But there is still ample separation between the plutonium

'-.a , peak and the Compton peak from any of the possible secondary targets to allow for a good

23
background measurement.

Tn Table I the last two columns show the relative plutonium fluorescence intensities and

the plutonium fraction of the total plutonium and uranium counts obtained, including interpolated values in parentheses, for the secondary targets we have been discussing. It is seen that for a given power input a ruthenium exciting source produces the highest plutoniu

count rate of the practicable sources. Figure 8 shows a typical fluorescence spectrum from a mixed plutonium-uranium oxide fuel

pellet in a titanium holder and excited by a rhodium x-ray target. The two prominent peaks near the center are the L a peaks of uranium and plutonium. Uranium has an line that

could be measured, but plutonium has no other useful L-line as its Lg„ peak is completely

overlapped by uranium Lg^ and the plutonium Lg^ fluorescence is not efficiently excited

by any of the targets from technetium to silver. Consequently, only the L a^ peaks of each

element are measured in the analysis.

INSTRUMENTAL METHODS OF INCREASING THE PU COUNT RATE

The principal limitation of an energy-dispersive system in this application is a badly

restricted count rate. The voltage signals from each x-ray photon are shaped and stretched

out in order to achieve good energy resolution, so they pass through the amplifier relative

slowly. There are only two peaks in the sample spectrum (see Figure 8) that we want to integrate and measure. But all of the rest of the peaks must also be absorbed by the detec

tor, and the signals from the entire spectrum from 0 to 40 keV must pass through the preamplifier and the amplifier before any part of that spectrum can be discarded. The LQ

. peaks

of uranium and plutonium together contain only about 37% of the total counts that must get

through the amplifier.

Optimizing the Exciting Radiation

One method of increasing the plutonium count rate is to optimize the choice of exciting

radiation, as was discussed in the preceding topic. For a pellet containing 33% as much

plutonium as uranium, a ruthenium secondary target yields a plutonium peak 39% as big as the uranium peak and raises the total uranium plus plutonium L

a ^ counts to 37% of the total

spectrum.

Dual Detectors

A second method is the use of multiple detectors. A second detector, complete with

amplifier and ADC were added to the spectrometer in order to double the count rate by

summing together the counts coming from both detector systems. Dual detectors looking at the sample pellet from different directions will also aid in averaging out the slight

particle size effects.

Fast ADC and Direct Memory Access

In a third attempt to increase the count rate we have provided an alternative route for

the signal through a very fast Tracor-Northern 200 MHz ADC. From here, instead of going

through any microprocessors or terminal boards, the signal will go directly to the computei

memory in four microseconds via a direct memory access (DMA) interface. However, as long

as the amplifier is the major bottleneck there will be little improvement in count rate by

use of direct memory access.

We plan to make two more changes that should increase the plutonium count rate by an additional factor of about four.

Monochromators to Isolate Pu and U Peaks

Molecular Data Corporation in Cleveland has developed a new kind of x-ray monochromator

which is very compact, utilizing a cylinder about one inch in diameter and several inches

long, which does not require a mechanical goniometer and, by using curved focusing crystal

does not attenuate the x-ray beam nearly as much as conventional crystal monochromators do,

These cylindrical monochromators can be designed to permit transmission of only the desirec

portion of the spectrum and to discard the unwanted counts before they get to the detector

and amplifier. As a result, the exciting power can then be increased and the count rate

in the two analytical peaks can be more than doubled without increasing the load on the

amplifier.

These cylindrical monochromators work well with a point source such as the target insic an x-ray tube. It is not yet known how well they can function with a larger diffuse source such as a secondary target. But we will know by August. Molecular Data, under contract

to the Hanford Engineering Laboratory, is currently experimenting with an x-ray source

24
..milar to ours to find out how well these monochromators will perform in the energy range

thorium, uranium, and plutonium L-^ lines with either a secondary target or with direct

icitation from a rhodium target x-ray tube.

i

1 Fast Amplifiers

The last modification will be to use faster amplifiers. Before the end of this year a

sw kind of integrating amplifier will become available. It does not stretch the signal

id can process counts in less than half the time needed by today's amplifiers, and the

isolution is said to be at least as good. Faster amplifiers may necessitate the use of

Lrect memory access. We shall find out.

If all of these techniques work out successfully we may achieve an eight- to ten- fold

icrease in plutonium and uranium count rates over our starting rate with a 1977 commercial

oectrometer. This means that a half million counts for plutonium could be accumulated in

3 to 70 live seconds and a relative precision (RSD) of ± 0.15% could be obtained for the

1 lutonium counts in two real minutes per pellet, including the time to change pellets.

opefully the overall analytical precision could be held to an RSD of + 0.20%. The accumulative average for 100 to 200 analyses per shift should provide an extremely accurate

lutonium inventory.

REDUNDANCY AND DOWN TIME

Figure 9 shows the entire x-ray spectrometer and glove box system as it was first

ompletely assembled the last week of 1977. It can be seen that there are two floppy

isc drives and two minicomputers. We did not plan instrument redundancy when we started

rocurement, but we have acquired a considerable amount of it. In the first place, we added a second detector system in order to double the count rate,

onsequently, if one of the detector systems should fail we could still carry on the pellet

\nalyses at reduced speed. At the time we purchased this equipment Kevex had no computer

»ut used a number of microprocessors controlled manually from the keyboard of the color

.isplay terminal. Therefore, we purchased a Tracor-Northern TN-11 system, which we were

amiliar with, in order to be able to program and control the entire operation automatically,

it the time we were to receive our spectrometer Kevex introduced their model 6100 processor

rhich includes a DEC LSI-11 minicomputer and a dual floppy disc drive, and they have

:ranslated most of their programs utilized by the microprocessors into FORTRAN for disc

iormatting. We exchanged a number of minor components and for an additional small relative

.ncrease in cost we were able to add the necessary interfacing and the new 6100 processor.

The second floppy disc drive was included because that was expected to have the highest

probability of failure.

Now we will be able to operate the entire system using either one of the disc drives

*ith either one of the minicomputers. Even if both computers should be down we still have

:he microprocessors that can be manually controlled from a keyboard. In most cases of

component failure this redundancy should prevent appreciable down time. At present the two ninicomputers use different languages. However, we have a FORTRAN compiler and a DEC RT-11

operating system in the Tracor-Northern part of the system and eventually we will convert

all our programs to a single language and format.

PRESENT CAPABILITIES AND FUTURE PLANS

There is considerable development work yet to be done, including the design of a completely automatic sample manipulator. When that and the programming are completed it will

be possible to control the entire system either locally or from a remote panel board. One

or the other minicomputer will automatically control the water flow and the x-ray generator,

choose one of several standards or sample pellets and control the sampling device, accomplish both energy calibrations and analytical calibrations when called for, measure the

radioactive background periodically and calculate a cumulative average for correction

purpose, acquire the analytical fluorescence data, calculate deconvolutions and matrix

corrections, integrate the corrected peak areas, calculate percent plutonium and percent

uranium, feed in the pellet weight, isotopic compositions, and oxygen-to-metal ratio

obtained from other close-coupled stations and calculate the percentages of plutonium and

uranium oxides and also calculate the absolute weights of plutonium and uranium, and

finally store the desired information on a disc and communicate the results to the master

control center.

If desired, the spectrometer may also periodically analyze a pellet for about 15 of the

impurity elements of interest by counting for a longer period of time. In a matrix this

dense these would include only those impurities heavier than chlorine and that have specifi25
cation limits of 200 ppm or more. It would be possible to have a second x-ray source and detector system in another glove box analyzing the oxide powders as soon as they are mixe

both for plutonium and uranium composition and for some of the impurities. This could be controlled by the same minicomputer that is controlling analyses of the sintered pellets

We plan for the next year or so to analyze every vendor fuel pellet selected for chare

erization before it is subsequently dissolved and assayed by amperimetric titration. Thj

will provide a firm fix on the calibration accuracy and also a comparison of the x-ray tl layer analyses with the complete dissolution and accountability determinations. We will

know then whether the pellets are as homogeneous as we think they are. By August we should have experimental data for the cylindrical monochromators . During

the following year we may be redesigning the x-ray source and detector compartment to make room for the monochromators. This would entail changing to direct excitation by th< x-ray tube, but the monochromators would remove essentially all of the background. We a. plan to incorporate faster amplifiers to permit faster counting and better precision. We

must also complete all the programming, design a completely automatic sampling device,

work out the bugs in the automatic equipment, and train some operators.

In its final design the automated x-ray spectrometer will be used in support of fuel

fabrication in the high performance fuel laboratory to be built as part of the Hanford

Engineering Development Laboratory.

Should the mixed oxide fuel pellets be changed in the future to carbide or nitride

fuels this would have absolutely no effect on the spectrometer system except to require new calibration coefficients.

If the fuel material should be changed to recycled Th-232/U-233 fuel, which will cont

a small amount of U-232 and decay daughters of Tl-208 and Bi-212 which emit gamma radiat

in the energy range 0.51 to 2.6meV,the increased shielding needed would require a redes

of the detector housing, the glove box and the sampler. The calculated flux of this hig

energy radiation is too great to permit use of Dr. Pickle's arrangment of two rate meter

one to measure total counts and the other to measure the high energy counts. In this ca the detector would not have sufficient time to recover between high energy photons. It fairly certain that crystal monochromators would be necessary to sufficiently reduce the

intensity of high energy radiation reaching the detector. This would not only lengthen

the distance between sample and detector, thus decreasing the x-ray intensity, but it wo cause a further large loss of x-ray intensity by insertion of the monochromators. Thus

analyzing times would be appreciably lengthened and precision decreased as compared to t analysis of uranium and plutonium oxide fuels.

26

Figure 2. Inside of the Detector and X-ray Source Compartment

28
Figure 3. Detector and Source Compartment Bolted to the Glove Box

29
Figure 4. Sample Compartment Separated from End Plate Inside Glove Box

30

Figure 6. Rhodium Exciting Spectra (50KV) Scattered from Polyethylene

32
Figure 7. Rhodium L X-rays Scattered from Different Matri

33
TABLE I

FLUORESCENCE YIELD VS SEPARATION BETWEEN

EXCITATION ENERGY AND ABSORPTION EDGES

EXCITASEPARAT,0N > keV NET PEAK AREAS, KC

TARGET X-U L... X-Pu L U L

a

Pu L

a

Mo 0.32 1400 220 0.136

Tc 1.25 0.34 1400 600 (0.30)

Ru 2.12 1.21 1400 563 0.287

Rh 3.05 2.14 1400 528 0.274

Pd 4.02 3.11 1400 494 0.261

Ag 5.00 4.09 1400 461 0.248

RATIO

Pu/(U + F

34
Figure 8. L-Fluorescence Spectrum from A PuC>2 UO2 Fuel Pellet

35
Figure 9. Complete X-ray Spectrometer System

36
Mechanisms for Assuring Accurate Measurement of Nuclear Materials

by

H. T. YOLKEN

National Bureau of Standards, Washington, D.C. 20234

ABSTRACT

A review of the procedures available to provide assurance of

meaningful measurements of nuclear materials is presented. These

procedures include the use of reference data, document or written

standards, reference materials and instrument calibration services.

Examples are given that demonstrate the determination of measurement

uncertainty relative to nationally accepted standards. A clear case

is made for the necessity of building a nuclear materials measurement

system on a firm foundation of accuracy.

KEYWORDS: Nuclear safeguards; measurements; standards

INTRODUCTION

Ihe determination of the amount of fissionable materials in nuclear fuel cycle

ities is a goal of both our domestic safeguards program and the international

ar safeguards activities as well. Although measurements play a key role in these

ities, we have not fully developed and utilized all the standardization procedures

able to insure accurate and compatible measurement results. Accuracy, in addition

ecision in the measurement of nuclear materials for accountability purposes, is

ly a necessity if one is to achieve meaningful results. However, measurements made

cessive accuracy levels are overly expensive financially and can be technically

iting. Therefore, the limits to measurement uncertainty, or in simple terms, the

acy of the measurements should not be grossly in excess of the level that is sufficient

et the end purpose. _

ACHIEVING ACCURATE MEASUREMENTS

Accurate and compatible measurements on a national and international scale have been

Lved utilizing several different means. For example, the measurement of mass for

I rcial transactions takes place smoothly since weights traceable through NBS to the

's primary standard are readily available to verify commercial scales. This assures

I tibility over both time and geographical location. Time measurements also enjoy an

i lent reputation for accuracy and compatibility. Time measurements, made both by the

nal Bureau of Standards (NBS) and the Naval Observatory, using an NBS-developed atomic

reference method, are constantly compared and coordinated with other countries' time

I rements. The results are then disseminated via radio broadcast. Since users accept

radio signals as their reference point, compatibility is achieved. Over the years,

e accuracy requirements for the measurements became more and more demanding, NBS was d to develop increasingly more sophisticated reference measurement methods. The

: d can now be determined to better than 1 part in 10 million, which is comparable to

' 1 second in 370,000 years.

Many other measurement systems draw on somewhat similar mechanisms to achieve uniform,

: ate results. For example, measurements for steel production, clinical chemistry health

the production of radiopharmaceuticals, and environmental monitoring all relate to a nally recognized system of standards provided jointly by NBS and others.

37
STANDARDIZATION FOR NUCLEAR SAFEGUARDS MEASUREMENTS

As the nuclear safeguards measurement system increased in sophistication and as lai numbers of research, instrumentation, regulatory and production organizations entered tl system, both domestically and internationally, the need for effective standardization hi increased. In order to help meet this need, NBS has recently embarked on a broad stan- dardization program for nuclear safeguards measurements. The NBS effort is supported b; the Nuclear Regulatory Commission (NRC) and the Department of Energy (DOE) . The under

taking is focussed on providing standardization for all of the measurements used for

material accountability purposes to all of the parties making the measurements. Commer

facilities, NRC regulators, government facilities, DOE inspectors, the International At Energy Agency (IAEA) and other countries all need, in general, to use an accurate,

consistent set of standards to make the system work effectively. Of course, in many ca these same standards will serve for production quality control, environmental monitorin

and buyer-seller equity in trade.

I should emphasize that NBS, as in many other measurement areas, is drawing upon t' expertise and technology developed by other organizations. We will also develop cooper

tive efforts with other organizations such as New Brunswick Laboratory (NBL) and Los Al Scientific Laboratory (LASL) , that have already made substantial contributions to the

standards effort. NBS is also strongly supporting the nuclear safeguards-related

standards-writing efforts of the American Society for Testing and Materials (ASTM) , the American National Standards Institute (ANSI) and the Institute for Nuclear Materials

Management (INMM).

The NBS program is addressing standardization of measurements in areas that incluc

bulk quantities (mass, volume, flow, pressure, density), destructive chemical and isotd

I--.-- J _•_ J__, — Amn J 1 ml.- analysis, and passive and active non-destructive assay (NDA) techniques. The statistic

aspects of materials accountability and sampling are also an important aspect of the el NBS is not aiming to develop field methods of measurement; our job is that of providinj

calibration methodology and data, standard reference materials (SRM's), calibration

services and where need be, developing reference measurement methods for characterizin

transfer standards and field methods.

NBS is also strongly committed to the concept of voluntary measurement assurance programs that allow participants to assess their measurement results relative to natioi

standards. These efforts also provide a two-way dialogue between the participants and

standards laboratory. The participant is given, where requested, guidance on how to improve measurement results and the standards laboratory gets valuable feedback on the

effectiveness of the standards-type services that they provide. Weaknesses in methodo

and instrumentation can also become apparent in measurement assurance programs.

New Brunswick Laboratory's Safeguards Analytical Evaluation (SALE) program is a good

example of an effective Measurement Assurance Program (MAP) for chemical analysis. NB

a cooperative project underway to develop a MAP for mass determination of large (up to* 17,000 kg) UF 6 cylinders. Plans to help initiate additional measurement assurance pro

for volume and various NDA techniques are also being formulated

If]

iatr

Sate m

to

(ill

lira

NDA REFERENCE MATERIALS

The NBS philosophy of how to provide for standardization of NDA measurements can be summarized by discussing the three areas of NBS activity in the NDA area. In the f' area, NBS is developing and certifying a small, select number of primary NDA SRM's for

materials that are in widespread use in one or more nuclear fuel cycles. An example i this area will be discussed shortly.

In the second area, we are developing reference measurement methods for the chara

terization of both NBS reference materials and "custom made" reference materials produ

by production and other types of facilities. These reference methods, which will most likely be costly and involve very large, expensive apparatus, will in general, not be

suitable for use as field methods. However, they could be used as an independent meas ment method to help validate field methods. Possible reference methods include: resonance neutron radiography currently under development at NBS using either large lj

38
.ctor-produced neutrons, Van de Graaff genereated 500 kev neutrons for interrogation

toto-nuclear reactions using large linac-produced gamma rays to include fission and

.etecting delayed neutrons.

j|in the third area, NBS is developing calibration techniques for NDA. Included in this

ire our activities to assist voluntary standards-writing organizations in developing

•n consensus standards for NDA measurements. We also plan on trying to develop

i red calibration techniques that might, in part, be based on nuclear data or radio-

: _ty reference materials and calculation methods. Of course, the hardest problem that

i to be addressed is how to detect and correct the mismatch between a reference material

i unknown sample, especially in inhomogeneous materials.

I

I would now like to describe a few ongoing projects at NBS that illustrate our approach

: mdardization of NDA methods. In the first example, NBS and the European Atomic

w Community (EURATOM)'s Ispra Laboratory agreed to try to develop and certify a series

\f enriched U3O8 Standard Reference Materials (SRM's) for gamma spectrometry. The

l Laboratory, with approval and guidance from the European Safeguards Research and

1 jpment Association (ESARDA) obtained the participation and support of several other

torn labs, while in the U.S., NBL and LASL agreed to join NBS in the effort. The IAEA

also provide an observer at key steps in the process. The result will be NBS-certimd distributed SRM's for the U.S., European Commission-certified and distributed

for the European Community and IAEA-certified SRM's for use by IAEA inspectors.

SRM's will be the first truly nationally and internationally-certified NDA SRM's.

rhe SRM's, which will be in the form of an infinite thickness for the emitted gamma

will be extensively characterized by chemistry and NDA. They should be useful in

.alibration of gamma-based NDA methods used for the measurement of U3O8, U0 2 , and UF 6. oriate how-to-use documentation will be provided with the SRM's. Upon completion of

low enriched U3O6 materials, NBS will most likely undertake (hopefully in a joint

jt), a series of high enriched U3O8 SRM's for calibrating NDA instruments.

: Ihe next NBS project I would like to discuss concerns the development and certifica-

: of plutonium heat source Standard Reference Materials to be used in the assay of

nium by calorimetry. The first step in this project was completed about a year ago

support by Mound Laboratory. The task involved a comparison of Mound heat flow

imetry results with NBS calorimetry results obtained by Ditmars 1 with an ice

imeter. Agreement between the two independent calorimetry methods was quite good. In urrent phase of thw work at NBS, heat flow calorimeter and associated constant

rature baths are being set up and intercompared with the ice calorimeter. This or

, well characterized heat flow calorimeters will then be used to measure and certify

nium heat source reference materials. Details of the certification program have not

een completely developed.

•These NDA activities are in agreement, in part, with the recommendations of a recent

Advisory Group on Physical Standards for NDA. The group recommended that the IAEA

the help of member countries) , certify a very limited series of primary NDA reference

ials. These reference materials should be very well characterized by both chemistry

DA measurements and will be applicable in a wide number of fuel cycle facilities.

materials recommended include uranium metal, plutonium metal and plutonium oxide (all

a range of isotopic compositions).

In addition, the IAEA Advisory Committee recommended that "custom made" calibration

ials for unique fuel cycle applications could usually be best made at the facility

haracterized and certified via independent NDA measurements. In some cases,

d A. Ditmars, "Measurement of Average Total Decay Power of Two Plutonium Heat Sources

Bunsen Ice Calorimeter," International Journal of Applied Radiation and Isotopes,

>, Vol. 27, pp. 469-490.

39
destructive chemical and isotopic analysis would also be needed. NBS does not plan, in

general, to provide this type of SRM. However, as discussed earlier, we do have an extensive effort to develop reference NDA characterization methods for reference materii

CHEMICAL AND ISOTOPIC REFERENCE MATERIALS

I would now like to discuss our efforts to develop and cerify additional SRM's for

destructive chemical and isotopic analysis. New SRM's that are planned or underway inc a high purity, thorium metal assay SRM, a Pu-244 mass spectrometry spike SRM, a U-233 m spectrometry spike SRM, a series of UFg isotopic SRM's for gas mass spectrometry and

ion-resin beads loaded with U and Pu for mass spectrometry. In addition, the isotopic

composition of three existing NBS plutonium isotopic SRM's have been remeasured and new certificates will be issued. The new certificates will permit individual users to calculate, on a day-to-day basis, the decay of the individual isotopes and the build-up

other elements due to this decay. Future plans also provide for the existing NBS pluto

isotopic SRM's to be certified by assuming that plutonium exhibits similar behavior to uranium on evaporation from the filament of a mass spectrometer.

Many of these SRM's for chemical and isotopic analysis will be cooperative undertakings; for example, Oak Ridge National Laboratory will supply materials, LASL will pi

purification and initial characterization and NBL will provide space and support in the plutonium-handling lab for a NBS plutonium chemistry lab to work on plutonium SRM's.

One of the most important facts built into most NBS SRM's is a known level of acci

The certification of the SRM's that were just described will be established by either c the following (in order of preference)

1. Use a reference method of analysis run independently by two or more

analysts (at least one of whom is at NBS)

2. Use two or more reliable independent methods (at NBS).

In the first case, a reference method (according to the NBS definition) is a meth

of known or proven accuracy, and its use assures the accuracy of the determination if

personal bias is eliminated. This is why the measurements must be made, if possible,

two or more analysts. An example of a reference method is the isotopic analysis of

uranium by isotopic dilution mass spectrometry.

The second way to certify an NBS SRM, used when a reference method (or absolute m

is not available, consists of using two or more independent methods. Each of these

independent methods must have estimated systematic biases that are small relative to accuracy goal set for certification. An example of this route to certification is ill

trated by the recently completed replacement uranium oxide SRM tor uranium assay. Thi

material was certified through a coopeartive effort with NBL. NBS obtained assay resu by coulometry and NBS and NBL obtained assay results by the NBL-developed Davies Gray

method. Agreement between the two methods was quite good.

CONCLUSION

The rapid advances in measurement methods , the need for increased and demonstrate

accuracy for safeguards purposes and the worldwide expansion of nuclear fuel cycle

facilities have all increased the need for NBS to strengthen its nuclear materials

standardization activities. The NBS program outlined is in response to this demand,

program is interactive by design and is drawing and building on the expertise and tect developed by the entire field.

40

ft

iilll
Design of a Cf-252 Neutron Assay System for Evaluation

at the Savannah River Plant Fuel Fabrication Facility

by

T. W. CRANE, S. BOURRET, G. W. ECCLESTON

H. 0. MENLOVE, and L. G. SPEIR

Los Alamos Scientific Laboratory, Los Alamos, New Mexico 87544

and

R. V. STUDLEY

E. I. duPont de Nemours and Co., Savannah River Plant

Aiken, South Carolina 29801

ABSTRACT

A nondestructive assay (NDA) unit will be evaluated at the Savannah

River Plant (SRP) reactor fuel fabrication facility for measurement of a range of highly enriched uranium materials. The unit employs cyclic

neutron interrogation with a Cf-252 neutron source followed by delayedneutron counting to assay the U-235 content of fuel alloys with up to 2.4 kg U-235 per item in addition to scrap and waste with a lower U-235

content. The accuracy goal for the majority of the measurements is 1-3%.

KEYWORDS: Nondestructive assay, enriched uranium, neutron interrogation,

Cf-252, U-235, delayed neutrons

INTRODUCTION

Test and Evaluation Program

As part of the Nuclear Safeguards Program at the Los Alamos Scientific Laboratory

ASL) several nondestructive assay (NDA) instruments are being evaluated or will be

'aluated in production facilities for monitoring the transfer and inventories of special

lclear material (SNM) . Use of NDA instruments should provide the capability for more

jnely accountability than is possible with destructive analysis methods alone.

In one of these test and evaluation programs a "Cf-252 Shuffler" 1 will be installed at \e Savannah River Plant (SRP) reactor fuel fabrication facility. LASL has been responsible

>r the design and construction of the unit within the constraints of the routine material

indling procedures at SRP, available floor space, and expected degree of operator profiLency. SRP will provide calibration standards corresponding to their material types and

irsonnel to operate the unit. LASL and SRP will share in documenting the results of the

ist and evaluation program.

Principle of Operation

The "Shuffler" measures the U-235 content of the SRP samples by counting the delayed

iutrons produced by irradiating the sample with a Cf-252 spontaneous fission neutron

)urce. The mechanical design of the "Shuffler" unit is shown schematically in Fig. 1.

ie unit consists of a source storage position and a sample interrogation-and-counting

ill. A motor is used to cyclically transfer the Cf-252 source between the shielded

zorage position and the sample irradiation position. In order to achieve a mechanically

Lmple system requiring few control features for assays, the SRP unit was designed to proLde a uniform response throughout the range of sample heights using only one interrogation

I. 0. Menlove and T. W. Crane, "A Cf-252 Based Nondestructive Assay System for Fissile

laterial," to be published in Nuclear Instruments and Methods.

41
position. In contrast, the first laboratory prototype "Shuffler" used an elevator to position the sample within its assay chamber, and two positions were selected by the

operator to obtain a uniform interrogation.

The assay begins with a background count with the sample in the interrogation well and

the Cf-252 source in its shielded storage position. Upon completing the background count,

the assay proceeds with neutron activation of the sample by transferring the source to the

irradiation position. Delayed-neutron counting follows after the source is transferred

back to its storage position. A cyclic sequence of irradiation and delayed-neutron counting continues until a sufficient amount of data has been accumulated or the maximum time

limit is reached. The final step in the assay is the calculation of the U-235 content of

the sample. The assay value is then available to the operator and the SRP accountancy

system.

SRP Materials

The primary material being handled at the SRP reactor fuel fabrication facility is

highly enriched uranium (up to 93% U-235). The uranium is alloyed with aluminum and then

fabricated into fuel elements. The materials to be assayed by the Shuffler include 1) pure

scrap from machining and other parts of the process that may be returned directly to production melts, 2) contaminated scrap and various products from the SRP scrap recovery

process, 3) contaminated scrap not recoverable at SRP that is shipped to Oak Ridge for

processing, and 4) waste materials without significant U-235 content that are sent to the

SRP burial ground.

The materials to be assayed will be packaged in steel cylinders (177.8 mm ID and 304.8 mm high). Standardizing the container size enables the assay to be accomplished more

simply with less chance of a container being improperly positioned within the sample assay

well. Large items requiring assay will be cut or broken up until they fit into the

standard container.

Material Flow and Measurement Points

In the fabrication of fuel elements of highly enriched U-Al alloy, more than 50% of

all alloy cast is recycled as scrap. About 90% of this scrap is suitable for direct

recycling into the casting process. This material is relatively clean with its composition

accurately known from process records. The remaining 10% of the scrap requires additional

processing before recycling. Most of this scrap is processed and recovered at SRP. The

low-level uranium waste product of the SRP recovery process is presently assayed with a gamma-ray NDA system. The items without significant uranium content are sent to the burial

ground.

About one per cent of all the U-Al scrap is not suitable for recovery at SRP and

includes floor sweepings, crucible pieces from the casting operations, and air filters. If the U-235 content is below that which is feasible to recover, the scrap is sent to the

burial ground. Scrap requiring reprocessing is packaged and shipped to Oak Ridge for

recovery. The "Shuffler" will measure most of these items before shipment.

SHUFFLER DESIGN FOR SRP

Meutronics Calculations

Some of the items to be measured by the "Shuffler" could contain as much as 2.4 kg of

U-235. With this amount of uranium self-shielding can be significant, especially for

thermal neutrons. In order to reduce self-shielding effects, the assay chamber is lined

with boral (B^C) and cadmium to remove low-energy and thermal neutrons while maintaining a penetrating (energetic) neutron flux. Ideally the interrogating neutrons should have

energies above several keV so that the resonance and thermal region is avoided and the

fission cross section is reasonably independent of the neutron energy. However, to limit

biasing contributions through fissions of the isotopes U-234, U-236, and U-238, the interrogation neutron energies should be below where their fission cross sections are appreciable, typically between about a half and one MeV. The Cf-252 source neutrons have a fission spectrum with the average neutron energy

being about 2.35 MeV. 2 To reduce the contribution from the nonfissile isotopes, the Cf-25

2 Reactor Physics Constants, ANL-5800 (July 1963).

42
eutrons are moderated. A series of Monte Carlo studies was made using the LASL MCN code^

o compare the relative effectiveness of various moderating materials for producing a enetrating interrogation with the least amount of fissioning of the U-234, U-236, and

-238 isotopes. Iron (steel) proved to be the most satisfactory choice for meeting the

Dove criteria when the additional factors of cost, availability, fabrication, and safety

ere considered. Neglecting the practical considerations, beryllium was the best choice

or lowering the contribution from isotopes other than U-235 and tungsten gave the most

anetrating spectrum. Neglecting cost and machinability considerations, nickel is

lightly preferred over steel.

A diagram of the assay chamber is shown in Fig. 2. This view is an enlargement of art of Fig. 1 with the material types noted. The moderating/reflecting material is

rimarily steel although some nickel is used between the source and the assay chamber. The

angsten piece surrounding the source at the interrogation position is primarily for

ersonnel gamma-ray shielding. The voids in the moderating assembly improve the uniformity

f the interrogation over the sample length.

The effects of self-shielding are partially compensated for by multiplication within

ne sample. For changes in the diameter of the alloy castings the effects of selfyielding and multiplication are nearly compensating. The data in Table I illustrate the

ffect; the smaller diameter item gives a higher induced-fission rate per gram because of

ass self-shielding but a lower delayed-neutron counting efficiency due to the lower

ultiplication. The relative response given in Table I is the product of the fission rate

nd the counting efficiency. The uniformity of the response for differing diameters is

xcellent and thus variations in the diameter of the ingots will not require a set of

tandards or correction factors based upon a separate measurement of the diameter.

TABLE I Data from the Monte-Carlo simulations of the Shuffler

designed for SRP. The quantity in parentheses is

the 1-0 uncertainty in the last digit given.

Relative Relative

Diameter fission rate/ Councint Relative

(cm) g Uranium5 Efficiency^1 Response

14.00 1.106(19) 0.988(13) 1.004(23)

15.90 1.004(17) 0.992(14) 0.996(22)

17.78 0.980(16) 1.019(14) 0.999(21)

aIsotopic composition: 231+U-1.2%, 235U-70.0%, 236 U-18.9%, 238U-9.9%

^Counting efficiency for delayed neutrons including multiplication effects

For cast ingots the mass is proportional to the height, resulting in the low-mass

amples being somewhat pancake shaped and the high-mass samples being tall cylinders. This

nange in the sample geometry with mass affects the response of the system. For the low-

: ass samples (pancake shaped) the response per gram rises because the neutrons can interroate the sample more readily due to the increased surface-to-volume ratio. On the other

'and, the response per gram for the largest samples tends to increase slightly because of

ultiplication. To compensate for both effects, the neutron-tailoring assembly (see Fig. 2) as designed so the interrogating neutron flux would be slightly lower at the top and

3ttom of the assay chamber. The flux depression was selected to cancel the increased

esponse of short samples at the bottom of the assay chamber and to offset the multiplicaion present in the heavy samples. Figure 3 shows the relative response of the system

ased on Monte Carlo calculations of the interrogation and delayed-neutron counting

fficiency. The data displayed indicate a linear esponse to within the accuracy of the

alculation (± 3%, 1 a).

E. D. Cashwell, J. R. Neergaard, W. M. Taylor, and G. D. Turner, "MCN: A Neutron Monte

Carlo Code," Los Alamos Scientific Laboratory report LA-4751 (January 1972).
Small variations from linearity could be included in the calibration function or in a correction factor. If the correction factor approach is used, then a measure of the sample

height can be obtained from the ratio of delayed neutrons detected by the bottom counters

to the number of delayed neutrons detected by the side counters. This ratio decreases as the sample height increases because the fraction of the delayed neutrons detected by the

side counters increases.

The accuracy of the measurements can be further improved by either correction factors

or internal consistency checks based upon additional data available at the time of assay.

Changes in the interrogation neutron flux, such as additional moderation or absorption in

the sample that affect the fission rate, are monitored by placing small neutron detectors

inside the assay chamber directly viewing the sample. Additional data to be obtained

include the sample mass measured by calibrated precision load cells located in the assay

chamber.

System Electronics

The electronics package contains the neutron detector power supplies, pulse amplifiers

and a microcomputer that records the data and controls the measurement sequence including

the source transfer, sample rotation, and load cell weighing. The detector preamplifiers

are built into the high-voltage junction boxes attached to the neutron detectors. This

location for the preamplifiers was chosen to minimize the noise pickup and thus improve th< reliability. The remainder of the electronics except for the interlocks and a control

console is contained in a temperature/humidity-controlled rack located just outside of the

sample assay room. The electronics are located outside of the assay room to avoid the

possibility of contamination. Should an item fail inside the assay room, it will be

replaced and the failed component will be disposed of.

A hardcopy output stating the NDA measurement of the U-235 content and its measuremen uncertainty as well as the total mass of the sample is given for each assay. This informa

tion is also available to the operator on a large-format video display screen visible

through a window in the assay room. A floppy disk storage unit provides a complete archi-j

val record for each assay. The detailed data stored on the disk file includes the

detector's counts from the He-3 neutron detectors and the flux monitors, temperature, line)

voltage, time of day and date, the load-cell data, and the sample identification.

System Software

The "Shuffler" software is being developed using Fortran IV rather than machinedependent assembly language. The software system will then be transferable to most

standard computers with a minimum of modifications. The Fortran coding will also facilitate the use of sophisticated algorithms for calibration and calculation of the U-235 mas:;; from the sample measurement.

The microcomputer will be programmed so that the details of the assay will be

completely handled by the computer. The operator needs only to enter the sample identifi

cation number and even this step could be eliminated if an automatic label reader is adde

to the system. The computer controls the assay sequence and performs all calculations ani

implements the correction factors based upon information obtained during the assay. Oncej

the assay is complete, the data is recorded as noted earlier and the computer informs the

operator that the system is ready to assay another sample. If a sample does not assay

consistent with the standards, the operator is informed and the inconsistency is recorded) J for later study. The computer will also inform the operator of minor problems such as neglecting to put the sample in the assay chamber or leaving the assay chamber door open, a

Standards and Calibration

During the initial phase of the test and evaluation of the "Shuffler" at SRP, the

system will be calibrated with test samples. These items will span the range of uranium

loading and be carefully prepared following a predetermined procedure. Once all the tesi samples have been analyzed by the "Shuffler," some of them will have their total uranium

content and isotopic composition precisely measured by destructive analysis. As the tes and evaluation proceeds, actual scrap and waste materials produced at SRP will be

destructively analyzed after having been measured. Any biases will be fully investigate I Eventually as experience is gained with the "Shuffler," a complete set of standards and

calibration procedure will be developed.

44
SUMMARY

The "Cf-252 Shuffler," a nondestructive assay (NDA) instrument, will be tested and

iluated for the measurement of highly enriched uranium at the SRP reactor fuel fabricaDn facility. The method of neutron interrogation with delayed-neutron detection is

Lieved to be the most practical solution for NDA of the enriched uranium materials found

i SRP. Pending a successful evaluation at SRP, the "Shuffler" unit will be used to litor the transfer of uranium alloys within the plant and to establish an assay value for

ste materials going to either disposal areas or recovery operations.

ACKNOWLEDGMENTS

We would like to thank Dr. S. C. T. McDowell and Dr. S. Baloga, Division of Safeguards

3 Security, Department of Energy, and Dr. R. B. Walton, LASL, for their efforts to Ltiate and coordinate this test and evaluation program.

|

45
S.R.P. SHUFFLER

DETECTORS

Figure 1. Schematic Diagram of the "Shuffler" Assay System for the

Savannah River Plant. A Sample is Shown in the Interrogation-Counting Well.

46
47
SAMPLE HEIGHT (cm)

0 5 10 15 20 25 30

0 0.5 1.0 1.5 2.0 2.5 3.0 3.5

URANIUM-235 MASS (kg)

Figure 3. Delayed-Neutron Response as a Function of the U-235 Mass

and Corresponding Sample Height.

48
Gamma Spectrometric Methods for Measuring Plutonium

R. GUNNINK

Lawrence Livermore Laboratory, University of California, Livermore, California 94550

ABSTRACT

Nondestructive analyses of plutonium can be made by detecting

and measuring the gamma rays emitted by a sample. Although qualitative and semiquantitative assays can be performed with relative

ease, only recently have methods been developed, using computer

analysis techniques, that provide quantitative results. This paper

reviews some new techniques developed for measuring plutonium. The

features of plutonium gamma-ray spectra are reviewed and some of the

computer methods used for spectrum analysis are discussed. The discussion includes a description of a powerful computer method of

unfolding complex peak multiplets that uses the standard linear leastsquares techniques of data analysis. This computer method is based

on the generation of response profiles for the isotopes composing a plutonium sample and requires a description of the peak positions,

relative intensities, and line shapes. The principles that plutonium

isotopic measurements are based on are also developed, followed by

illustrations of the measurement procedures as applied to the quantitative analysis of plutonium liquid and solid samples.

KEYWORDS: Plutonium analysis; gamma-ray analysis; computer analysis;

plutonium safeguards.

INTRODUCTION

The accountability and safeguarding of plutonium as a special nuclear material

ntinues to be an important issue. Material balance and control require that an accurate

ventory be kept wherever plutonium is processed or handled. As a result, it is necessary

have methods for accurately measuring plutonium in its various forms. Not only must the

tal elemental amount be determined, but the amounts of the individual isotopes from mass 8 to 242 must usually also be measured. An assay of 241Am is also frequently of interest.

The procedures for making these measurements have traditionally involved a combination

I two or more destructive chemical methods. Although these procedures have become highly

xurate and will undoubtedly continue to be a mainstay of the analytical laboratory, they

ive the disadvantages of requiring considerable time and effort and of being difficult to iplement in-line. For these reasons, nondestructive methods have been sought that might

rovide a more timely and more automated analysis.

We have been developing techniques for nondestructive analysis methods using passive

ounting techniques by detecting and analyzing gamma rays emitted by or transmitted through

plutonium sample. In the following discussions, we will develop some of the bases for

hese analyses, described some applications, and present typical results we have obtained.

AWork performed under the auspices of the U.S. Department of Energy by the Lawrence

ivermore Laboratory under contract number W-7405-Eng-48

49
PLUTONIUM SPECTRA

Measurement Considerations

Taken with a germanium detector, the spectra of plutonium samples are very complex as Fig. 1 shows. All of the isotopes except 242pu make detectable contributions to the spectrum, but to varying extents and at different energies. The gamma-ray energies and absolute emission rates have been carefully studied and published in previous reports. 1" From a review of these data, one can deduce several important characteristics that are useful for spectral analysis:

239 • Pu exhibits significantly intense peaks over the energy region from 50 to 800 keV.

238 2A0

• The Pu and Pu isotopes emit relatively few gamma rays, with the most inter ones being of low energy.

• Only about seven or eight energy regions (see Fig. 1) exhibit contribution from more than one isotope. 3,

• The energy spacing in some of these regions is such that the peaks overlap sevei

10fc

10'

§ io3 u

10

100

Energy — keV

200 300 400

1 if

100-keV

9rouP

, 148-keV 125-keV group

group '

1 ^

160-keV

group

208-keV

group 335-keV 370-keV

group group

630-keV

group

1-kg of weapons-grade Pu02

66 min at 0.25 keV/ch

500 600 700 800

Energy — keV

900 10

lli,

Fig. 1. A typical spectrum of aged plutonium. The intensified peak regions contain

contributions from more than one isotope.

50
These characteristics obviously present some problems for spectrometric analysis.

:e analyzing them further, we note that gamma-spectrometric analyses are also dependent

le physical characteristics of the sample, such as:

• Physical/chemical form (e.g. solution vs solid)

• Age of the material since chemical processing

• Concentration of the plutonium in the sample

• Homogeneity of the sample

• Elements or isotopes contained in the sample

• Isotopic grade of the plutonium

The combination of these physical factors with the gammas-emission characteristics of

.plutonium isotopes makes it impossible to have a single method or technique that can be

to analyze all sample types. Instead, one must tailor the method to the type of

jrement. For this reason, it is important to establish a foundation of basic data and

analysis methodology that, when properly developed and understood, can be modified and

according to the prevailing circumstances.

Spectral Analysis Techniques

Although the plutonium gamma -ray spectra are quite complex, some of the isotopes

to it single, well-resolved peaks. The amount Xk of such an isotope in a sample can be

I rmined by integrating to find the net area of a given peak, Aj , and dividing by the

Ling efficiency E

j 5 k- Tnat is

A. =J(y.-b.),

y

i

= y. counts in channel i, i

background counts in channel i.

ore than one gamma ray of a given isotope or if more than one isotope contributes to same peak, a more generalized expression can be written as follows

A

j = 1 I e:,k x

k

' (2)

k=l j=l

As long as the equations do not form an ill-conditioned set, they can readily be

'ed by the method of linear least squares.

Some of the plutonium isotopes, however, can only be determined from spectral regions

happen to be quite complex. To resolve the overlapping peak multiplets in these

.ons, we must first have a description of the peak shape. Although the peaks in

lanium-detector spectra are principally Gaussian in shape, some tailing does occur,

:icularly on the low-energy side. Therefore, the algorithms that we have formulated to

the data in a peak region, such as shown in Fig. 2, usually include a central Gaussian

>onent, a background continuum, a short-term tailing and sometimes a longer-term tailing,

51
10*

10fc

c

©

10«

10J

1 1

I r 105 keV

FWHM = 500 eV

Gaussiandistribution

Ge

escape peak

/ /

Background

continuum

Short-term

Long-term

tailing

J/

tailing

-J

/

// ~>

I

I

20 40 60 SO 100

i \!

\i

II

i\ j_LVi

120 140

Channel

Fig. 2. Gamma-ray peak shape showing the Gaussian and

tailing portions of the response.

We have found that after subtraction of the background continuum, the following algorithm

adequately describes the component elements of a peak:

y. = y° exp [a(x

± - x

0

) 2

] + T^), (3)

where

y. = net counts in x.,

i i y^ = peak height at the peak position x^,

a = peak-width parameter,

and

T(x^) = tailing function.

The tailing function is given by:

2 T(x) = [A exp (Bx) + C exp (Dx)] [1 - exp (0.4ax )]<S , (4)

where A exp (Bx) accounts for the short-term tailing and C exp (Dx) accounts for the Ion term tailing. Also, A and C are tailing-amplitude parameters, B and D give the slope of

the tailing, and x = xi - x^. The final term involving a reduces the effect of T(x) to

52
• at the peak position, and 6, which has a value of 1 or 0, removes the effect of T(x)

positive values of x.

Although Eq. (3) is useful for describing gamma-ray peak shapes, it does not

mately describe the observed distribution of K-series x rays associated with the heavy

ents.5 Because of the short lifetime of K-shell vacancies, the Lorentzian energy disution of the emitted x rays has a FWHM of about 100 eV. When this distribution is conlted with the instrumental dispersion, the resulting peak shape, as shown in Fig. 3, is

stantially different from that of an equivalent-energy gamma ray.

The data values, Y., in an overlapping peak multiplet can be considered to be a linear

Ition of the appropriate contribution from each peak j so that

n n Y.=7 y. .= 7 y? • f(x., shape parameters). (5)

j=l j-1

Techniques have been developed for characterizing many of the shape parameters used

;he above algorithms. ^ However, these techniques were first used in generalized computer

grams, which usually also assume that the peak positions x^ and the peak amplitudes y^

always unknown and therefore must be considered as free, rather than fixed, parameters,

se codes must therefore treat Eq. (5) as a set of nonlinear equations and resort to some

rative method to solve them for the unknown parameters.

106

105

cJ 104

10-

1 I 1 1-

r 94 keV Lorenzian x-ray-v 1 1 *ev

energy distribution N

A

I

III

Instrumental

line shape

Convoluted - distribution

x V/ ft

v

102_[_ 20 30 40

Channel

50 60

Fig. 3. Detail showing the convolution of the intrinsic

x-ray distribution with the instrumental line shape.

53
However, for a spectrum resulting from a pure plutonium sample, both the exact ga ray energies and the relative intensities can be predetermined for each of the isotopes

interest using tables that report the gamma emissions of plutonium and its daughter

products. ^ Therefore, once an energy scaling has been established, it is possible to I late the exact position x^ where each of the peaks is expected to appear in a spectrum.

This means that all of the parameters in the exponentials of Eqs. (3) and (4) can be pri determined and therefore held fixed in the fitting process. The resulting set of equat

expressed by Eq. (5) then become linear in form and can be solved by a standard leastsquares method.

The number of free parameters can be further reduced by realizing that many of tb

peaks in a complex grouping might belong to the same isotope. Therefore, the relative

intensities y9 for each isotopic component k in the group can be predetermined and the

composite spectrum profile can be expressed by:

= y y y k=l j=l

(6

This means that component response spectra can be generated from known peak positions ai relative intensities, thereby reducing the number of unknown parameters to the number o: isotopic components. This procedure is illustrated in Figs. 4 and 5. Figure 4 shows tl 94- to 104-keV response spectrum that can be generated from the gamma and x-ray peaks

resulting from the 239pu component. Response profiles can be similarly calculated for i of the isotopes that contribute to this region, as shown in Fig. 5. When these spectra

properly normalized, they can be fitted by linear least squares to the observed composit

spectrum to yield the amounts of the component they represent.

10e

105

c

oo

10«

10-=

- I

1

I

1

I

1

I

1

I

\ ^98.44 keV

' I :

„r94.66 keV ."/ .*/ (x-ray) •(/ •A*

•/ \* *

* / i \* * *

•/

•/ (x-ray)

\\./\ *

• V *. :

• •

• •

• • • •

. •• • -

\\/~

h

98.78 keV

J u * • A

•/ 1* * / •/ l* • J

/ V

/ l\i

/ H"u r103.02

/ keV *

•/ r* ** /

v*

i/ v_/*^\ ^ yj

V

V\ \

{ - / \/V96.15keV |

94 96 98 100 102 104

Energy — keV

239

Fig. 4. The response spectrum of Pu in the 94- to 104-keV region. This spectrum can be calculated from

known energies and intensities of the five x rays and

gamma rays contributing to this region.

54
Fig. 5. The 94- to 104-keV region showing the response

spectra of the six plutonium isotopes contributing to this

region. Each isotope is represented by one or more radiations, as illustrated in Fig. 4.

One additional feature should be noted here. Equations corresponding to integrated

le peaks, as expressed in Eq. (2), can be added to the set of equations expressed by

(6), provided the equations are appropriately normalized and weighted. Figure 6 illuses the constitution of a matrix of coefficients for such a mixed set of equations.

A complete and quantitative measurement of the isotopes requires that the efficiency

ficients and normalization factors be carefully calibrated. Frequently, however, it is

necessary to determine the ratios of the plutonium isotopes. Such ratios can be easily

ulated without extensive calibrations, provided that for each pair of isotopes two or gamma rays of similar energy, but arising from the different isotopes, can be found and

rately measured. For example, the isotopic ratio A/B can be given by:

1^ t. /I t_\-l I. e

9 P 0 t ePVeP; I e P t 1 1 \ 2 2/ 2 1 1 B

1^ = peak intensities of the gamma rays arising from isotopes A and B, respectively,

G l'

e 2 = count in8 efficiencies for the respective gamma rays,

P^, ?2 ~ the gamma yields or branching probabilities of the respective gamma rays,

55
c

u

c0

a,cOua

o

241 Am

241 Pu

240pu

" 239pu

238 Pu

Isotopic response spectra Efficiency coefficients rmFTi

A.

TT

A

_L_L

95 104 59 94 129 148 152 208

Energy — keV

140 equations-

(using net channel counts

for Y in Eq. (6) )

6 equations

(using net peak counts

for A in Eq. (2) )

and

Fig. 6. Illustration of a heterogeneous set of equations to be

solved for unknown isotopic abundances. Some of the equations

are set equal to the net peak counts while others are equated

to the net channel counts.

"A' B

t = the respective half-lives for A and B.

Although it might not be possible to obtain accurate values of and P-^ individua]

it is possible to obtain accurate ratios over a small energy range. The half -lives are well known, and therefore the accuracy of the calculated isotopic ratio is largely limite

by the precision with which the peak intensities 1^ can be measured.

"

ha Isotopic ratios can be calculated from a grouping of resolved peaks by multiplying, ^ ejjk in Eq. (2) by Pj,k« A program called RATIO was written to calculate isotopic ratios)

" using data from the 94- to 104-keV region. This program first computes response spectra lt each of the isotopic unknowns using peak-shape information, library branching intensities

the detector-efficiency curve, and some estimated physical characteristics of the sample u its containment. Because the efficiency ratios (e^/e^) can only be estimated, an additici

parameter is allowed to be free. The coefficients of this unknown are simply the channel

counts multiplied by the channel value. When this profile is included in the fitting

process, the effects of small residual efficiency disparities are effectively removed. l\ .: effects of this process are shown in Fig. 7.

56
Channel

Fig. 7. Using the additional response function (a) in the

fitting process lowers the residual efficiency disparities.

Part (b) shows the residual spectrum without the additional

function; (c) shows the residual spectrum when the

additional function is included.

There are about eight peak groupings from which isotopic information can be obtained,

se have been studied in some detail using computer simulation techniques.^ Because most

:he groupings do not contain peaks from all the plutonium isotopes, it is useful to

Dine several groups so that all of the ratios can be accurately established. To do this,

start by modifying Eq. (2) so that it takes the following form: m n Vll \ > (8)

k=l j=l

re E = counting efficiency for the group.

57
To obtain the isotopic ratios with respect to one component, we can write Eq. (8) a n m n

(9)

Table I. Precision of plutonium isotopic-ratio measurements by gamma-ray spectrometry f reactor-grade plutonium solutions.

Ml

in Vi j=l k=2 j=l

ilutio If we let R^ — Xj^/X^ and generalize the above equation using s different peak groupings,

|

er7j obtain

n s m n 2

p3,i- I rvH pja\ • (10: j=l 1=1 * k=2 j=l

From this equation, we see that one additional degree of freedom is required for e| ^

new grouping that is added. The set of equations can be put partially or totally into a response-spectrum format, as discussed earlier, rather than in the peak-area form shown : Eqs. (8)-(10).

APPLICATION AND RESULTS OF THESE TECHNIQUES

As we have indicated, the application and expected results of gamma spectrometric

methods cannot be divorced from a consideration of the physical and chemical characteris

of the sample. A first convenient division is to consider plutonium to be either in a solution or in a solid.

Plutonium in Solution

Plutonium in solution form is handled in many places and will be the principal fori in reprocessing facilities. The design of counting cells and the method of counting and

spectral analysis depends on such factors as time since chemical processing occurred, th

concentration of the solution, the isotopic grade, and the presence of foreign activitie

Recently processed solutions containing "^Pu, 239pu _ 240pu> an(j 241pu exhibit intense p at low energies. As a result, relatively short counting periods ^10-2U min) can provide

accurate results. Some typical results are shown in Table 1.

Solutions that have aged several days since chemical purification rapidly grow in

content and therefore require a different procedure of spectral analysis. The computer

analysis program is somewhat more complex, and a longer counting period is generally

required. Typical analysis results for such samples are also shown in Table I.

Precision (la) , % Abundance, Recently separated Aged

Isotope wt% solution Solution

238

Pu 0.230 0.46 0.44

239

Pu 76.06 0.07 0.07

240

Pu 19.09 0.26 0.26

241

Pu 2.98 0.45 0.23

241

Am -0.17 - 0.38

aValues obtained from ten or more replicate analyses.

58
Solution isotopic analyses of weapons-grade plutonium samples are continuously made

the Savannah River Plant and at Lawrence Livermore Laboratory by counting 10-ml portions

(a special counting vial. Also, we have performed several experiments at the Allied

.eral Nuclear Services plant*** 9 to demonstrate the feasibility of using gamma spectrometric

i hods for in-line measurement. A dual-cell system, schematically represented in Fig. 8,

r designed to permit both direct counting of the gamma rays emitted by the isotopes in

i ution and a total plutonium-concentration measurement. The latter is performed by

i erving the differential attenuation of two transmitted gamma rays, one on either side of

K-shell absorption edge of plutonium. Figure 9 shows an assembly of the experimental

i aratus that is now being tested. Although the experiments are not yet completed, the

i liminary results are encouraging.^*^

Plutonium in Solids

The analysis of plutonium in solid samples by gamma spectrometry presents additional

I ficulties. Some problems here are controlling or defining the counting geometry, severe f-attenuation of the gamma rays by the sample, and potential sample inhomogeneity.

Fig. 8. Schematic drawing of the experimental arrangement used in a gamma-ray

absorptimeter . In the top drawing, the Ge detector records only the natural

radioactive gamma rays emitted by the plutonium solution from the thin front

cell; this data yields the plutonium isotopic distribution. In the bottom,

the Ge detector records predominantly the highly collimated ^5ge an(j 57qq gamma

rays that pass through the solution in the long transmission cell; this data

is used to determine the plutonium concentration.

59
Fig. 9. Photograph of experimental set-up illustrated by Fig. 8. The set-up simulates in-line measurement of plutonium streams. The arrangement is designed

to accommodate both a direct measurement of the solution and a K-edge differentialattenuation measurement. The transmission source is not in place.

One analysis method that overcomes most of these problems combines gamma-ray spectrometry with calorimetry. The former provides isotopic-ratio data and the latter measures

total power production of the sample. Because both measurements are nondestructive, quanti

tative analyses can be made without breaching the containment of the sample.

This method has been implemented at Lawrence Livermore Laboratory for the routine

assay of 1- to 2-kg lots of plutonium oxide whose isotopic content is about 93% J7Pu. We

obtain the low-energy gamma spectrum by rotating the container in front of a small, l-cm^

high-resolution germanium detector, as shown in Fig. 10. We use the program RATIO,

mentioned earlier, to calculate the isotopic ratios from data found in the 94- to 104-keV

region of the spectrum (see Fig. 5). The program generates component response spectra for

2 -^Pu, 239pu> 240pU5 241pu> 241^^ for a ipha-induced plutonium x rays, and for small

efficiency disparities over the region being fitted (see Fig. 7).

Table II compares results produced by RATIO with mass-spectrometry results. Although

counting periods of several hundred minutes are generally used, reasonably good precisions

can be attained with much shorter counting periods, as Table III indicates.

The procedure, used in RATIO, of analyzing only the 94- to 104-keV group is adequate

for weapons-grade samples, where the 239pu content is high. To obtain sufficiently accurat

isotopic ratios for reactor-grade materials, we will have to expand the program, using

Eq. (10), so that more groups will be simultaneously considered. We might also need a two- detector system. We will still need a small but very high-resolution detector for the 94-

to 104-keV region, but we will also need a larger, more efficient detector to obtain a sufficient counting rate in the 300- to 400-keV region.

60
-Plutonium

oxide sample

-Collimator

Fig. 10. Simple arrangement for measuring the plutonium

isotopic ratios of plutonium-oxide samples.

Table II. Standard deviation between gamma-spectrometry isotopic-ratio results and mass spectrometry results.

Relative standard deviation, % Approximate

Isotope abundance, wt% Solutions3 Solidsb 238

Pu 0.01 2.5 2.5

239

Pu 93.5 0.03 0.03

240

Pu 6.0 0.33 0.62

Pu 0.5 0.88 1.3

0.01-0.2 2.7 4.0

Comparison of 24 analyses.

Comparison of 7 analyses.

61
Table III. Precision of plutonium isotopic-ratio measurements using the RATIO program for' analyzing weapons-grade plutonium oxide.

Precision (lo) , %, a for various counting periot

Isotope Abundance, wt% 4 min 40 min Ton f 12 U mm 600 mi

5.1 3.0 2.0 0.7

0.17 0.05 0.03 0.013

2.3 0.9 0.5 0.2

1.2 0.6 0.2 0.1

2.5 0.9 0.4 0.4

238. 239

240

241

241

Pu

Pu

PuPu

Am

0.01

93.7

5.8

0.44

0.1

Values obtained from ten or more replicate analyses.

CONCLUSIONS

We have established a good foundation of basic spectrometric data for the gamma emissions and decay properties of the plutonium isotopes and their attendant daughter products

We have also made progress in developing techniques and algorithms for analyzing the

spectral data and in forming a methodology that allows the analysis programs to be tailored

to the type of sample and measurement that is to be made. Although additional developmenta

work will be required, we are now at a point where gamma spectrometric techniques can be

used for the routine measurement of plutonium in several forms.

FOOTNOTES

1 R. Gunnink and R. J. Morrow, Gamma-Ray Energies and Absolute Branching Intensities for

238,239,240,241pu and 241^, Lawrence Livermore Laboratory, Rept. UCRL-51087 (1971).

2 J. C. Cline, Gamma Rays Emitted by the Fissionable Nuclides and Associated Isotopes, Idahc

National Engineering Laboratory, Idaho Falls. Idaho, Repts. IN-1448 Q970) and ANCR-1069

(1972).

^R. Gunnink, J. E. Evans, and A. L. Prindle, A Reevaluation of the Gamma-Ray Energies and

Absolute Branching Intensities of 237tj

> 238 , 239, 240, 241pu> anc[ 241^^ Lawrence Livermore

Laboratory, Rept. UCRL-52139 (1976).

*R. Gunnink, A Simulation Study of Plutonium Gamma Ray Groupings for Isotopic Ratio

Determinations, Lawrence Livermore Laboratory, Rept. UCRL-51604 (1974).

?R. Gunnink, Nucl. Inst. Methods 143, 145 (1977).

R. Gunnink and J. B. Niday, Computerized Quantitative Analysis by Gamma-Ray Spectrometry,

Lawrence Livermore Laboratory, Rept. UCRL-51061, vol. I (1971).

7 R. Gunnink and J. E. Evans, In-Line Measurement of Total and Isotopic Plutonium Concentrations by Gamma Ray Spectrometry, Lawrence Livermore Laboratory, Rept. UCRL-52220

(1977).

62
Reference to a company or product name does not imply approval or recommendation of the

product by the University of California or the U.S. Department of Energy to the exclusion

of others that may be suitable.

I1 K. F. Hofstetter, G. A. Huff, R. Gunnink, J. E. Evans, and A. L. Prindle, "On-Line

Measurement of Total and Isotopic Plutonium Concentrations by Gamma-Ray Spectrometry,"

paper presented at the 21st Annual Conference on Analytical Chemistry in Energy Technology,

October 4-6, 1977, Gatlinburg, Tennessee.

63
Rapid Nondestructive Plutonium Isotopic Analysis*

by

J. E. Fager and F. P. Brauer

Battelle, Pacific Northwest Laboratories, Richland, Washington 99352

ABSTRACT

Methods for plutonium isotopic measurements have been evaluated for

nuclear safeguards inventory verification. A mobile, real-time, non- destructive assay, gamma-ray spectrometric measurement system has been

assembled, moved and operated at several nuclear storage facilities to perform rapid, real-time plutonium isotopic measurements on Pu metal,

Pu nitrate and Pu oxide.

KEYWORDS: Nondestructive analysis (NDA) ; plutonium isotopes;

gamma-ray spectrometry; x-ray spectrometry

INTRODUCTION

At Battelle, Pacific Northwest Laboratories methods for performing plutonium isotopic

measurements have been evaluated for use in verifying nuclear material safeguards inventories. Since effective verification requires the on-site analysis of a significant fraction of the total inventory in a relatively short time period, rapid analytical methods

are necessary. Methods of calculating plutonium isotopes by gamma-ray and x-ray spectrometry have been reported by several investigators, 1-5 but these methods have previously

been used primarily for the nondestructive analysis (NDA) of laboratory nuclear material

samples in which analysis times could be long and precise results were required. Nuclear

material inspectors need to make equivalent measurements at nuclear storage facilities but are limited by time and the large number of items to be measured. To meet this need,

plutonium isotopic measurements could be used along with other NDA measurements, such as neutron counting and gross weight, to verify the nuclear material content of items selected

by the inspectors. These measurements need to be made at the nuclear material storage

facility since it is impractical to move large quantities of nuclear material to the labora

tory to make verification measurements. This paper describes rapid plutonium isotopic

measurement techniques using gamma-ray and x-ray spectrometry that can be used for on-site

nuclear inventory verification.

J. L. Parker and T. D. Reilly, Plutonium Isotopic Determination by Gamma-ray Spectroscopy . LA-5675-PR, Los Alamos Scientific Laboratory, Los Alamos, NM, pp. 11-19 (1974).

2 R. Gunnink, J. B. Niday and P. D. Siemens, A System for Plutonium Analysis by Gamma-Ray

Spectrometry, Part 1: Techniques for Analysis of Solutions . UCRL-51577, Lawrence

Livermore Laboratory, Livermore, CA, (1974).

3 R. Gunnink, "Status of Plutonium Isotopic Measurements by Gamma-ray Spectrometry."

Proceedings of American Nuclear Society Meeting , New Orleans, LA, 1975.

^J. F. Lemming, Gamma-ray Spectroscopy Development . MLM-2485, Mound Laboratory,

Miamisburg, OH, pp. 8-11 (1977).

5 F. X. Hass and J. F. Lemming, "Plutonium Isotopic Measurements by Gamma-ray Spectroscopy."

Proceedings ERDA X- and Gamma-Ray Symposium , Ann Arbor, MI, 1976.

6 F. X. Haas and J. F. Lemming, Plutonium Isotopic Measurements by Gamma-ray Spectroscopy . MLM-2286, Mound Laboratory, Miamisburg, OH, pp. 5-15 (1975).

*This paper is based on work sponsored by the Department of Energy under Contract

EY-76-C-06-1830.

64
EQUIPMENT DESCRIPTION

high-resolution gamma-ray spectrometer interfaced to a minicomputer was used for

experiments. The spectrometer is part of a mobile, real-time nondestructive assay

systam, which with data processing and data storage equipment, is mounted in a illy constructed vehicle. The nuclear material detection equipment is transported in

| hicle and is then moved into the nuclear material vault area to perform the measureC while the data processing and storage equipment, a multichannel analyzer and analogd ital converted (ADC), remain in the vehicle. Interfaced to the computer, the analyzer

controlled remotely from a portable terminal located in the nuclear material storage

a ith the detector. This system has been described in other reports.

he measurements were made with a planar germanium detector with a 500 mm2 active area

1 mm thick and a resolution of 560 eV for 122 keV gamma-rays. For some of the measure- i the detector was operated 70 m away from a multichannel analyzer and ADC, and the

i .onal cable between the ADC and detector amplifier reduced the resolution at 122 keV

f • eV. This resolution loss did not significantly interfere with the plutonium isotopic

* -ements. The detector was operated with a gain of 0.075 keV/channel and the count rate

. ;pt at 5000 counts per sec or less. The low-energy detector response was reduced by

n lg absorbers consisting of 0.02 cm of copper and 0.15 cm of cadmium between the

I cor and the sample.

\ coaxial germanium detector 43 mm long with a resolution of 980 eV at 122 keV was used

i isure the higher energy gamma-ray spectrum. When this detector was operated with 50 m

| ole between the ADC and detector amplifier, the 122 keV resolution dropped from 960 eV

: 30 eV.

Ihe computer system located in the mobile vehicle consists of a PDP-11, 32K word

Iter, dual RK05 disk drives, terminal and electrostatic printer. The system is also

[ faced with the NDA equipment. Gunnink developed the data analysis technique we

. ted to evaluate for rapid plutonium analysis. ~ 10 Originally, the computer code for

i ata analysis was written in Fortran II to run on a 12-bit, 28K word minicomputer

; faced to a disk storage device. For these measurements, it was modified to run on the

' 1 16-bit, 32K word computer using Fortran IV and the RT-11 operating system supplied

gital Equipment Corporation (DEC) . The Fortran language and system library routine

i ied by DEC were used for all program modifications. Since dual disk drives were

I able on our system, we had ample disk storage for the operating system, programs,

|: tor calculation data, raw spectral data and calculated results.

; Two of Gunnink 's data analysis methods were evaluated. 11 * 12 The first method generates

;ponse function for each of the components of the 94 to 104 keV peak multiple from

mium, americium and alpha-induced x-rays. The response functions are fit by a leastres method and the plutonium isotopic composition is calculated. This method requires

the sample matrix and material between sample and the detector be well known.

?. Brauer, J. E. Fager, J. H. Kaye and R. J. Sorenson, "A Mobile Nondestructive Assay

ification and Measurement System," J. of Inst, of Nucl. Mat'ls Mgmt. Ill, pp. 680-6

77).

aink, A System for Plutonium Analysis by Gamma-ray Spectrometry , UCRL-51577.

nink, "Status of Plutonium Isotopic Measurements by Gamma-ray Spectrometry."

Gunnink and J . E . Evans , In-line Measurement of Total and Isotopic Plutonium Concentrans by Gamma-ray Spectrometry . UCRL-52220, Lawrence Livermore Laboratory, Livermore, CA,

4.nink, A System for Plutonium Analysis by Gamma-ray Spectrometry , UCRL-51577.

nink, In-line Measurements of Total and Isotopic Plutonium Concentrations by Gamma-ray

ctrometry , UCRL-52220.

65
The second method calculates the plutonium isotopics from ratios of selected

adjacent gamma-ray peak areas. The computer code developed by Gunnink will calculate

selected peak areas and unfold selected peak multiples by describing the fundamental peak

shapes. If the ratios listed in Table I are measured and the mass fraction of 238Pu +

239Pu +

240Pu +

2ttl Pu = 1, the four equations with four unknowns can be solved to yield

the plutonium isotopic abundance. The computer code to automatically calculate the

selected peak areas and plutonium isotopic ratios and reduce these to plutonium isotopic

abundance has been written.

TABLE I . Peak Pairs Used for Pu Isotopic Calculations

from Gamma-ray Spectra

Nuclide Gamma Energies Factors to Convert

Ratio (keV) Peak Area-ratios to Atom Ratios

238Pu /2tfl Pu 152.8/148.6 1.183

239Pu /2«*lpu 208.0/203.5 1585.

2t*0pu / 2i+1 Pu 160.3/164.6 50.7

The computer time required to run either program is from 2 to 4 minutes, depending on how many iterations are required for the least squares fit to converge to the

desired precision.

EXPERIMENTS

Several investigators have reported branching intensities and gamma-ray energies fo: 237U, 238Pu, 239Pu, 21t0 Pu, 2kl ?u and 21+1 Am. 13 Since 2kzTu has no gamma-rays associated

with its decay, it cannot be measured by gamma-ray spectrometry. Table I lists the pluti

isotopic ratios and the gamma-ray energies of the peaks used in the isotopic calculation:

Table I also lists the factors used to convert the peak area ratios to plutonium atom

ratios. These factors are the gamma-ray intensity ratio divided by the ratio of the

nuclide decay constants. This method of plutonium isotopic computation was suggested by

Parker and Reilly, 11+ but the gamma-ray intensities and decay constants are from the work

by Gunnink. 1

5

Tables II through X summarize the various experiments performed to compare the

plutonium isotopic computation methods. The National Bureau of Standards (NBS) SRM-946

plutonium standard was counted 9 times for 200 seconds and the plutonium isotopic abunda was calculated from the 100 keV spectral region and by the gamma-ray peak pair method

(Table II). The large uncertainty in the 2 ^°Pu estimate is caused by the 2ltl Pu interfer

at 160 keV peak. The gamma-ray branching intensity ratio for the 159.9/164.6 keV peaks

from 21+1 Pu is 0.1488. l ° Since the 160.28 keV Zh0Tu peak cannot be resolved from the 159

21+1 Pu peak, the 164.6 keV 21+1 Pu peak area must be used to correct the 160 keV peak area

for the 2hl ?u interference. For the NBS SRM-946 4.17% 2kl?u standard, about 70% of the

peak at 160 keV is due to the 21+1 Pu interference in the 2lt0 Pu measurement. The high

standard deviation of 2.39 for the 21+0 Pu estimate reflects the uncertainty in the 160.28

240 Pu area estimate.

13 R. Gunnink, J. E. Evans and A. L. Prindle, A Reevaluation of the Gamma-ray Energies an Absolute Branching Intensities of 237U, 238 > Zi3 > 21*" » zttl Pu, and 21tl Am . UCRL-52139,

Lawrence Livermore Laboratory, Livermore, CA, 1976.

11+ Parker, Plutonium Isotopic Determination by Gamma-ray Spectroscopy . LA-5675-PR.

15 Gunnink, A Reevaluation of the Gamma-ray Energies and Absolute Branching Intensities

of 23 7Ut

T3-B, ZTTj zmj, zni, Pu and zm^ UCRL-52139.

16 Ibid.

66
TABLE II . NDA Plutonium Isotopics of NBS SRM-946

Derived from Gamma-ray Peak Pair Ratios

Weight Percent

200-Second Count 2 3 8pu

2 39pu 240pu 241pu

X U . iJ7 8A ^fi XUi n .DO£a *t . UD

9 nU 9i 7 • ZX / 87 Sfi 7 1Q "5 Q7

"1J n 9 ^n OO . Oy a r>7 f . u / 4 0.209 83.80 11.24 4.08

D nU • ZZ99Q so. in J

5 . O8TJAH . 98ZO

c0 n 9^i 8OO5 • ISXO A4 . 19XZ

7 n 969 81OX 7fi • / D XZ1 9 • DO56 A4 • 8AO'f Q0 u

n . Z9i£jg 8T0 88 J . OO XX11 9 A . L OA f\K H . UD

Q n 9 SAOH Afi . 4D XUi n .DOfis a nfi

Mean 0.232 85.49 9.55 4.17

Std. Dev. 0.015 2.50 2.39 0.26

NBS Value 0.246 83.05 12.11 4.02

Difference 0.014 -2.44 2.56 -0.15

% Difference 5.6 -2.9 21. -3.7

TABLE III . NDA Plutonium Isotopics of NBS SRM-946

Derived from 100 keV X-ray and Gamma-ray

Spectral Region

Weight Percent

%

(-Second Count 23 8Pu 23$ Pu 24C Pu 241 Pu

1 0. 275 83. 55 12. 21 3. 97

2 0. 258 83. 44 12. 36 3. 94

3 0. 229 84. 58 11. 38 3. 81

4 0. 253 84. 29 11. 60 3. 86

5 0. 236 84. 78 11. 17 3. 81

6 0. 250 85. 49 10. 57 3. 69

7 0. 233 83. 95 12. 06 3. 76

8 0. 253 83. 57 12. 12 4. 06

9 0. 236 84. 75 11. 25 3. 76

Mean 0. 247 84. 27 11. 63 3. 85

Std. Dev. 0. 015 0. 70 0. 60 0. 12

NBS Value 0. 246 83. 05 12. 11 4. 02

Difference -0. 001 -1. 22 0. 48 0. 17 Difference -0. 4 -1. 5 4. 4. 2

TABLE IV . NDA Plutonium Isotopics of NBS SRM-948

Derived from Gamma-ray Peak Pair Ratios

Weight Percent

200-Second Count 23 8Pu 239 Pu 240 Pu 24 x Pu

1 0. 011 92. 29 7. 19 0. 473

2 0. 010 91. 81 7. 66 0. 484

3 0. 013 92. 51 6. 98 0. 463

4 0. 009 92. 63 6. 85 0. 476

Mean 0. 011 92. 31 7. 17 0. 474

Std. Dev. 0. 002 0. 36 0. 36 0. 007

NBS Value 0. 011 91. 54 7. 94 0. 472

Difference 0. 0 -0. 77 0. 77 -0. 002

% Difference 0. -0. 8 9. 7 -0. 4
i'

TABLE V . NDA Plutonium Isotopics of NBS SRM-948

Derived from 100 keV X-ray and Gamma-ray

Spectral Region

Weight Percent

Second Count 2 38pu

239Pu 2k °Pu 2k x Pu

1 0.0194 92. 24 7. 30 0. 442

2 0.0106 91. 92 7. 62 0. 456

3 0.0227 92. 05 7. 47 0. 456

4 0.0104 91. 95 7. 59 0. 450

Mean 0.0158 92.04 7. 50 0. 451

Std. Dev. 0.0062 0.14 0. 15 0 007

NBS Value 0.011 91.45 7. 94 0 472

Difference -0.0048 -0.5 0. 44 0 021

Difference -44. -0.5 5. 5 4. 4

TABLE VI . NDA Plutonium Isotopics of NBS SRM-948 Derived From 100 keV X-ray

and Gamma-ray Spectral Region as Function of Counting Time

Weight Percent

Counting Time (Sec) 2 38Pu 2 3 9pu 240pu 30 0.013 ± 0.003 92. 75 ± 0.56 6. 80 ± 0.56

60 0.026 ± 0.002 92. 73 ± 0.37 6. 80 ± 0.39

200 0.010 ± 0.001 91. 95 ± 0.18 7. 59 ± 0.19

500 0.013 ± 0.001 91. 89 ± 0.12 7. 64 ± 0.11

1000 0.016 ± 0.001 91. 47 ± 0.14 8. 05 ± 0.14

NBS Value 0.011 91. 54 7. 94

241 Pu 0.437 ± o.o; 0.446 ± o.o: 0.450 ± 0.0(

0.451 ± 0.0(

0.464 ± 0.0(

0.472

TABLE VII . Comparison of Facility and NDA Plutonium Isotopics for

34.4 Grams of PUO2 Derived From Gamma-ray Peak Pair Ratios

Weight Percent

Counting Time (Sec) 239Pu 2tt0 Pu 2ttl Pu

60 90.5 ± 1.2 8.8 ± 1.2 0.63 ± 0.012

100 92.8 ± 0.7 6.5 ± 0.7 0.63 ± 0.009

300 91.8 ± 0.4 7.6 ± 0.4 0.62 ± 0.005

Facility Value 91.5 7.9 0.59

TABLE VIII . Comparison of Facility and NDA Plutonium Isotopics for

66 Grams of PUO2 Derived From Gamma-ray Peak Pair Ratios

Weight Percent

68

Counting Time (Sec) 239Pu 2t*°Pu 2ttl Pu

60 89.5 ± 1.9 8.4 ± 1.9 1.9 ± 0.07

100 87.8 ± 1.5 10.1 ± 1.5 1.9 ± 0.05

300 87.0 ± 1.2 10.8 ± 1.2 2.0 ± 0.04

Facility Value 86.3 11.8 1.7
TABLE IX . Comparison of Facility and NDA Plutonium Isotopics

for Plutonium Metal Derived From Gamma-ray Peak

Pair Ratios Counting for 200 Seconds

of

239pu Weight Percent

240-11 Weight Percent 2UPu

Weight Percent

Facility NDA- Difference Facili ty NDA Difference Facility

3.05

1.85 1.98

0.66

NDA Difference

L

3

1

73.2 80.6

85.5

91.6

75.2 ± 2.9 81.4 ± 2.7 87.3 ± 2.4 94.8 ± 1.7-2.0

-0.8

-1.8 -3.2

22.6 17.1

12.1

7.6

20.4 ± 2.9 16.2 ± 2.7 10.5 ± 2.4 4.5 ± 1.7

2.2 0.9 1.6

3.1

3.34 ± 0.2 -0.29

1.94 ± 0.11 -0.09

1.92 ± 0.11 0.06 0.64 ± 0.03 0.02

TABLE X . Comparison of Facility and NDA Plutonium Isotopics

for Plutonium Nitrate Derived from Gamma-ray Peak

Pair Ratios Counting for 1000 Seconds

239pu 240p u 241p u

,»f Weight Percent Weight Percent Weight Percent Facility NDA Difference Facility NDA Difference Facility NDA Difference

87.2 85.7 ± 2.2 1.5 11.1 12.8 ± 2.2 -1.7 1.5 1.3 ± 0.1 0.2 90.8 90.5 ± 0.7 0.3 8.3 8.8 ± 0.7 -0.4 0.80 0.66 ± 0.02 0.14

90.8 91.0 ± 0.7 -0.2 8.3 8.3 ± 0.7 0.0 0.80 0.65 ± 0.02 0.15

Table III summarizes the plutonium isotopic abundance derived from the 100 keV spectral

)ns. In each case the counting time was 200 seconds and the low energy detector response

reduced by the addition of 0.02 cm copper and 0.15 cm cadmium absorbers between the detecand the sample. The difference between the NDA and NBS values for all four of the

rmium isotopes is less than two times the standard deviation of the nine observations.

Tables IV and V summarize the plutonium isotopic abundances calculated for the NBS

948 standard, Table IV by the peak area ratio method and Table V by the 100 keV method,

e the 24 *Pu abundance was 0.472%, the 160 keV peak area is subject to less interference

the 241 Pu. Lower uncertainty and better NDA-NBS agreement for the 2 ^°Pu estimate was

rved for the measurement of NBS SRM-948 than for the measurement of NBS SRM-946.

The high count rate of the 100 keV spectral region suggests that counting periods even

ter than 200 sec could be used to obtain plutonium isotopic information. Table VI lists

calculated isotopic abundances as a function of analysis time from 30 sec to 1000 sec. shorter counting period results appear to be biased low for the 24 ^Pu and 24 *Pu estimates

high for the 239Pu and 238Pu estimates.

Two larger Pu02 samples were analyzed by the peak area ratio method. Table VII

ares the facility value with the NDA value as a function of counting time for a 34.4 g de of Pu02- Table VIII summarizes the same comparison for a 66 g Pu02 sample. Four

onium metal samples of about 200 g were measured by the peak area ratio method. Table

:ompares the facility and NDA plutonium isotopic abundances calculated from a 200 sec

kt. Three 2 kg plutonium nitrate samples were analyzed by the peak area ratio method.

.e X compares the facility and the NDA results. A coaxial germanium detector was used

these measurements. The side-pointing 41 mm diameter by 43 mm-long detector has a Dlution of 980 eV for 122 keV gamma-rays and 1.93 keV for 1332 keV gamma-rays.

The errors reported in Table VII through X are at the la level and are derived from the

3t squares fit of the peak shape parameters. The agreement between the facility and NDA

jes for the plutonium oxide and nitrate measurements is better than 2a. Most of the

jits agree within la. These results were very encouraging and showed that a wide variety

materials could be measured at the storage facility and in a short analysis period.

69
CONCLUSIONS

The results to date indicate that the plutonium isotopic measurements can be performed

by gamma-ray and x-ray spectrometry in counting periods as short as 200 sec. Because of

the high count rate of the 100 keV spectral region, the plutonium isotopics could be

obtained in even shorter counting periods, but the sample matrix, sample container and

other material between the sample and the detector must be known to make use of this regioi

Most nuclear material in a storage facility is packaged in several metal containers, plast:

bags and bottles and facility records do not always show the complete packaging history.

This limits the usefulness of the 100 keV spectral region.

The higher energy adjacent gamma-rays peak areas used to calculate the plutonium

isotopes by the peak pair ratio method are less affected by sample matrix and packaging.

The 2 ^ 1 Pu concentration limits the precision in measuring the 2 ^°Pu 160 keV peak in short

counting periods. The 642.41 keV 2 ^°Pu gamma-ray peak has been used to estimate the 21+0Pu concentration. 17 No 24 *Pu gamma-ray peak is adjacent to the 642.48 keV 2t+0 Pu gamma-ray

peak, so detector efficiency and sample attenuation factors would have to be considered to measure the 2t*°Pu/ 2lf *Pu ratio by using the 642.48 keV peak. The decreased detection

efficiency for the higher energy gamma-rays also limits the detected counts at 642 keV in

short counting periods. Use of the 642.48 keV gamma-ray with the present detector systems

needs to be evaluated.

These plutonium isotopic measurements have been made at several nuclear material stor age facilities. The analytical results from these measurements are available to the inspe tor in less than 5 minutes and the isotopic results can be used along with other NDA

measurements to make real-time decisions about the contents of selected plutonium samples.

Thus, if necessary, the inspector could require additional samples to be analyzed or some samples to be reanalyzed, based on the real-time data analysis this system provides.

ACKNOWLEDGMENTS

The support of A. C. Walker of the DOE, Richland Operations Office and of

R. J. Sorenson, H. L. Henry and C. 0. Harvey of Bat telle is gratefully acknowledged. The

electronic assistance supplied by W. A. Mitzlaff is also gratefully acknowledged.

17Haas, Plutonium Isotopic Measurements by Gamma-ray Spectroscopy , MLM-2286.

70
Non-Destructive Assay of Leached Hulls in a Nuclear Fuel Reprocessing Plant

by

K. J. HOFSTETTER, B. C. HENDERSON, J. H. GRAY and G. A. HUFF

Allied-General Nuclear Services, Barnwell, South Carolina 29812

ABSTRACT

In a nuclear fuel reprocessing plant, the spent light water

reactor fuel elements are chopped into approximately 2-4 inch

lengths and the irradiated fuel dissolved from the Zircaloy hulls

in a nitric acid dissolver solution. From the dissolver, the

empty hulls are transferred batchwise to a leached hull monitor

to be assayed for undissolved uranium and plutonium. After the

assay, the hulls may be recycled back to the nitric solution for

continued dissolution or disposed of as solid nuclear waste using

the assay results to provide accountability data for special

nuclear material content.

The hull monitor at the Barnwell Nuclear Fuel Plant (BNFP)

will be a remotely controlled, fully automated system designed

to quantitatively assay leached hulls for undissolved uranium

and plutonium. The hull monitor will assay the hulls from one metric ton of fuel per dissolver basket with the design goal of

detecting 0.1% undissolved fuel and yet remain within the framework of the BNFP materials flow, i.e. , handle five hull baskets

per day.

The non-destructive assay will be accomplished using a com- puter-based gamma-ray pulse height analysis system employing a 5x5 inch Nal(Tl) scintillation detector. The intense radiations

from the fission product isotopes and the activation product

isotopes produced in the reactor prevent direct assay of the

undissolved fuel left in the hulls. The measurement will be

made indirectly by demonstrating a correlation between the

amount of ^^Ce undissolved and the remaining uranium. The

isotope "^Ce is a direct fission product with high (6%) cumulative yield. The daughter isotope lutf Pr has a gamma ray at 2.18

MeV well above other predominant radiations in the spectrum from

the major interferences 60 Co, 58 Co, 95 Zr( 95 Nb), 137 Cs and

l06 Ru(1o6 Rh).

Segmented scanning operation of the hull monitor is accomplished by rotation and vertical transversal of the hulls con- tainer past the detector station. Proper collimation and absorbers are required to maximize the ll4l*Ce(

11<l, Pr) to background

ratio. A basket indexer is provided which monitors the scanning

rate and ensures repositioning.

The leached hull monitor system will be interfaced to a computer-based multichannel analyser for ease of operation and

data handling. A calibration basket has been fabricated to accomodate radioactive sources and inactive Zircaloy hulls.

Experiments to be performed with the hull monitor using this

basket will simulate operation prior to plant startup. The

system, calibration, geometry and hull attenuation, spectral

interferences, quantitative limits of assay, and the role of

the hull monitor in nuclear materials accountability will be

discussed

.

KEYWORDS: Design of experiments; leached hull monitor; non- destructiye assay; evaluation, calibration

71
INTRODUCTION

The Leached Hull Monitor (LHM) was built by Intelcom Rad Tech for the Barnwell Nuclear . Fuel Plant and has been described by Gozani. 1 Its purpose is to detect the presence of un-'jL, dissolved fuel in chopped hulls of irradiated fuel elements after leaching with nitric acid The assay can be used to determine any shipper-receiver differences as well as for accounts*; fte bility of special nuclear materials. For a reprocessing plant with a capacity of 1500 metri jtlvei tons of spent fuel per year throughput, the leached hull monitor would detect and account fqjy £ approximately 4-5 kg of undissolved uranium per day. The position that has been applied to Lye the BNFP states that a quantity of fuel remaining in the leached hulls exceeding 0.1% of tb'Let input must be detected and accounted for. This corresponds to approximately 1 kg of uraniums

in each basket of hulls (about 650 kg of hulls are associated with one metric ton of fuel) .

|j, 0 f

The hull monitoring system involves passive gamma-ray techniques to scan the dissolver^ as basket for 1 4 k Ce (

1 4 "*Pr) , a fission product isotope produced at a 6% cumulative fission yielijL 0 f The daughter isotope 1!tl*Pr has a 2.186 MeV gamma ray which is relatively free of spectral

interferences. The ratio of lu

''Ce to uranium in the dissolver solution, determined by laM

e oratory analysis is used to calculate the amount of undissolved fuel remaining in the leached^

hulls. Jjito DESCRIPTION l-L

oped m 0

The hull monitor is an attempt to develop an improved non-destructive assay method foi measuring the undissolved fuel remaining in a basket of hulls from a chop-leach type nudes

fuel reprocessing plant. It has evolved into a specialized and sophisticated system as po- tential problems became apparent. Because of the extremely high radiation environment, th< detector, absorbers, and collimator are located in a sealed maintenance gallery. The elec- tronics, data acquisition and analysis system, and remote controls are located in the ana- lytical laboratory, a cable run of about 650 feet. The detector, absorber filters, and th< collimator are mounted on carts that can be independently moved to a position of optimum JJigtl geometry. The carts ride on rails bolted to the sides of the maintenance gallery. The de

tector can be placed a maximum of 12.5 feet from the hull basket. A schematic of the LHM

shown in Figure 1 .

; ^

ttati

1st

icess

To optimize the detector for highest efficiency at 2.186 MeV, a 5-inch by 5-inch Nal('

scintillation detector was chosen. The 2.186 MeV gamma ray is only 0.74% abundant in the

decay of '"'''Pr, and, therefore, maximum sensitivity is required. With this detector appro:

imately 82% photointeraction of the gamma ray occurs. The mean stopping distance of the

gamma ray in Nal is 1.75 in. or at 35% total depth of the crystal. The Nal(Tl) detector disi

plays excellent intrinsic efficiency for the 2.186 MeV gamma ray. A 4% full-width at half-,

( ,;ffe maximum resolution is observed for the 2.186 MeV peak using this detector.

The filter assembly is located between the collimator and the detector. It is cartmounted and remotely controlled for ease of operation. The purpose of the filter assembly

is to optimize the

11>l*Ce to background ratio. Three lead filters (1-inch, 2-inch, and 4-in

thickness) can be remotely placed between the detector and the radiation source. Because

different burnup and decay times of the fuel assemblies to be reprocessed, it may be neces

sary to change the thickness of the absorbers to minimize the intensity of low energy radi

tions and yet maintain reasonable yield for the high energy gamma ray. A pneumatic contro

system with electric solenoid switches was designed and installed on the filter cart assemb

to position the filters in the beam line. The hull monitor cannot be approached once fuel' assembly assay has begun due to the high radiation background. In fact, the entire hull

monitor system was designed under the criterion that maintenance and calibration would occ only during plant shutdown or two weeks out each year.

The collimator is a three-foot long bored lead cylinder with a 3-inch diameter collimating hole. There is a 3-foot by 3-inch by 0.180-inch wall copper sleeve down the center

of the bore to minimize X-ray and secondary radiation originating from the gamma-ray inter

action with the lead. The hull monitor is separated from the hull basket by a 16-inch wal

72

fete

i4s II

lot

|l to

Wet
.gh density concrete that has a 24-inch wide by 6-inch high window. This window has been

1 with lead brick to produce a 4-inch by 6-inch hole to further collimate the the gamma

itions. The collimator, concrete wall and the lead-lined window allows the detector to

'•5 on a relatively small portion of the hull basket as it is rotated and lifted past the

»w.

J The hull basket receives the hulls from the shear equipment and is lowered into the

olver solution. After dissolution of the fuel from the chopped hulls, the basket is

ad, allowed to drain, and lowered onto the hull monitor table. The table is rotated and

2d vertically while scanning is accomplished. The hull basket is approximately 3 feet

iameter and 7 feet high, and capable of holding the hulls from one metric ton of fuel.

= is a basket positioning transducer that permits repositioning of the basket for further

y of localized lumps of undissolved fuel. This basket indexing mechanism was installed

ne drive axle gear and is interfaced to the data acquisition system. The transducer

as as a monitor device for the consistency of the scanning rate and provides the capaty of repositioning the basket for futher study of irregular scans.

j The electronics and the data acquisition system are located in the laboratory where

rarature and humidity control can be closely maintained. The electronic system was deed to take advantage of pulse pile-up rejection techniques. The leached hull monitor

acquisition and analysis system (LHM-DAS) is a 16K PDP 11/05 computer with input/output

f mass storage devices. The LHM-DAS will be interfaced to the materials accountability

r uter for permanent record storage and status alert for plant operators. The system can

I rogrammed to locate hot spots and apply appropriate correction factors. The system was

r loped to automatically interpret data and calculate the quantity of undissolved fuel with

rmum operator intervention.

HULL MONITOR CALIBRATION

i

1 Specific calibration tests of the leached hull monitor detection system will be performed

[ ng the cold checkout period for performance evaluation of BNFP. The calibration tests

i he LHM will require a number of radioactive sources to duplicate the fission products

I

induced activity in the cladding, end pieces, and spacers present in the leached hull

.et during a normal basket scan. The sources will represent several concentrations of

ion products and cladding activity corresponding to 0.1% - 0.001% undissolved uranium

. remaining with the Zircaloy hulls. Unirradiated Zircaloy hulls will be placed in the

l .bration basket along with the radioactive sources using remote handling techniques.

Kblishment of the relationships between spectral data and radial position of the sources

i lecessary in order to determine the average mass of Zircaloy hulls between the source

the detector. Attenuation correction factors can be determined. Isotopes with possible

trference potential can be placed in the calibration basket in the same ratio found in

hulls to determine limits of assay.

The calibration basket is a stainless steel drum that will simulate the dissolver basket

.

is designed to hold calibration sources in thin tubes running vertically in the basket,

tubes are located at varying distances from the center of the basket so that different

:ix combinations of hulls and sources can be obtained. The calibration basket is filled

'i unirradiated Zircaloy hulls and the sources placed into the source tubes.

A small calibration check source of 1 "^Ce

(

1 4

u

Pr) is located on the detector cart that

1 be remotely moved into position directly in front of the detector. The check source

jl be used to verify the performance of the electronics and the detector response to

^e( 1 '*'*Pr) between calibrations with the calibration basket. It will be used to monitor

actor efficiency and amplifier gain shifts. The assay system is also equipped with a ser to monitor the detector preamplifier electronics.

Cladding and dissolver residue studies give some indication of the environment and

ntitative limits of assay expected for the leached hull monitor. 2 ' 3 Gamma activity can

ginate from several sources when scanning the basket of hulls and an understanding of

ir associations with the cladding will minimize the error for the total assay.

During the dissolution cycle, insoluble residue is formed which would be expected to nsfer out of the dissolver with the dissolver solution. Approximately 0.2 - 0.3 weight

73
percent of the irradiated U0 2 in light water reactor fuel is insoluble. The cladding can become coated with a thin film of this residue during dissolution in HN0 3 that is not neces sarily associated with undissolved fuel. Filtration and/or centrifugation of the dissolve

solution will collect a substantial black deposit with a weight of 0.05% to 0.3% of the tot* weight of the U02 depending on the type fuel and the amount of burnup; the lower the burnu|

the less residue is present. The dominant fission products are

103 Ru, 1 0

Ru( 1 0Rh)

,

llt0 Ba( llt0 La) 95 Zr( 95 Nb), 137 Cs, 134 Cs, 141 Ce and 1 * "Ce (

1 "

4 Pr) with the ruthenium isotopes i accounting for as much as 40% by weight. Very small amounts of uranium can be detected in the residue. The residue also contains Cs, Ce, Cm, and Pu activities in approximately the same ratio to uranium as in the fuel and therefore can be associated with undissolved fuel

The residue is extremely radioactive due to the fission product ruthenium. This will

require special attention because of the possible spectral interference with the 1 '*'*Ce(

ll*'t

; peak and its heat generating properties. Temperature gradients can cause performance vari

ations in the detector.

Gamma activity associated with the cladding is the result of the recoil of the fissio

products of nuclear reactions ejected from the fuel into the cladding as well as induced

activity in the cladding. The fission products have a much greater recoil energy than the transuranium elements. Thus the relative fraction of fission products trapped in the clad

ding would be much greater than the transuranium elements. Approximately 0.1% of the tota fission product inventory is found associated with the cladding. The fraction of transuranic elements associated with the cladding is 0.001 - 0.002% of the total present. 2 These

contamination levels will then contribute a minimum base or background in the hull assay

and will vary according to total burnup time and location in the fuel assembly. The recoi

of 141*Ce into the cladding is approximately 0.09% of the total in the fuel and unrelated t the amount of plutonium or uranium left undissolved.

Another potential problem in the non-destructive assay technique of leached hulls is

the intense gamma activities of 60 Co due to the activation of the end pieces and spacers

in the fuel assembly. The Zircaloy is a relatively pure alloy which contains very little

cobalt. Unfortunately, activation products in the core hardware greatly diminish the

spectral sensitivity and selectivity of the lltl*Pr peak. These end pieces and spacers con- tain up to six hundred times more cobalt as in the Zircaloy. Cobalt activation in the mis

cellaneous components could produce from 100 to 400 times the 60 Co activity observed in tl cladding alone. The 1 "^Ce

(

1 h 4 Pr) peak at 2.186 MeV is well above that for 60 Co; however,

400 fold addition of 60 Co to the spectrum makes the

lltl*Pr peak limits difficult to define.

This interference limits the quantitative capabilities of the monitor to perhaps 1-2% of

the total fission product inventory. It is anticipated that the electronic pulse pile-up

rejection system will make the spectral sensitivity to

lltl*Pr acceptable. Figure 2 is a spectrum taken of check sources of 60 Co and 1 "^CeC 1 ^Pr) in a gross activity ratio of 375

to 1. This system is capable of maintaining approximately 4% resolution with only 9-12%

system dead time.

These tests are indicative of the environments and quantitative limits expected of th« hull monitor. The spectral interference of 106 Ru( 106 Rh) as a dissolver residue, the recoi

of fission products into the cladding not associated with the undissolved uranium, and th< intense induced activity of the cladding hardware will be some of the limiting factors to the system. Time also becomes a limiting factor for the sensitivity of the leached hull

monitor as the spent fuel waits to be reprocessed. The decay of the 1 4Ce( 1 k "*Pr) (half

life 285 days) causes the ratio of 14

''Ce to longer lived fission products to increase.

It is estimated that spent fuel older than seven years can not be assayed using the lhh

'Pr 2.185 MeV gamma ray to quantify the amount of undissolved uranium or plutonium left in th< hulls. In that case, another gamma ray relatively free from spectral interference may be

chosen or an alternate method used to determine the amount of undissolved uranium.

References

l T. Gozani, "Leached Hull Monitor for Detection of Undissolved Irradiated Nuclear Fuel",

INMM, Vol. Ill, No. Ill, (Fall 1974).

2 B. L. Vondra and D. E. Campbell, "LWR Fuel Reprocessing and Recycle Program Quality

Report for Period January 1 to March 31, 1977", 0RNL/TM5864, p. 2-9.

74
vannah

56.

re 1.

gure 2.

River Laboratories Quarterly Report for ERDA LWR Programs", (December 1976),

LIST OF FIGURES

A schematic of the leached hull monitor system.

A portion of the spectrum of 60 Co and ^CeC 1 ^Pr) using pile-up rejection

electronics at a gross activity ratio of 350:1.

75
FIGURE 1

SHIELD

WALL

DRiVE MOTOR

76
FIGURE 2

< H 1 1 1 1 1- 4- 1 h

10,000-

_J

u

z

z

<

x

o

ccLJ

{2Z'3

O

8,000-

6,000-

4,000-

4-

2 ,000- + ++

++++ +

i r 60 80

t i i i i I I r 100 120

CHANNEL NUMBER

140 160

77
Safeguards and Nonproliferation

by

Samuel C. T. McDowell

Department of Energy, Washington, D. C.

ABSTRACT

U. S. nonproliferation objectives cannot be achieved without

the development and implementation of modern safeguards systems

that are timely, sensitive and perhaps most important, inspectable

by the International Atomic Energy Agency. In the nonproliferation

context, the purpose of safeguards is to deter or reveal national

intentions to obtain nuclear weapons capability by providing the

time required to enact international sanctions. The Department of

Energy is developing safeguards concepts and systems based on minimizing the opportunity for diversion and maximizing the probability of timely detection to allow for the necessary communication

and control actions. The current U. S. nonproliferation policy

can afford the time to make available practical, active safeguards

for assuring the world that nuclear materials are being used only

for peaceful purposes.

KEYWORDS: Safeguards and nonproliferation, active and passive

fuel cycle safeguards concepts , integrated safeguards

systems.

INTRODUCTION

The most critical issue facing the international nuclear community today is

nonproliferation. The term nonproliferation can be simply defined as the prevention

of weapons capability in a non-weapons state. For nuclear power to be considered a viable option for meeting future energy requirements, questions of proliferation,

safeguards and waste management must be resolved. The U. S. and general world agreement on a delay in large scale commercial reprocessing and recycle does, however,

allow some time to address the proliferation and safeguards issues in a meaningful

and systematic fashion. Waste management, I will leave as the subject for another

paper at another time.

Nonproliferation objectives cannot be achieved without the development and application of modern safeguards systems that are timely, sensitive, and perhaps most

important, inspectable by the International Atomic Energy Agency. While other major

considerations are also involved, including such things as types of fuel cycles,

coordinated sanctions, and declared national intent, it is only through the implementation of timely, reliable safeguards systems and adequate inspection technology and

procedures that we can be assured strategic quantities of weapons material have not

been diverted from nuclear facilities. In the nonproliferation context, the purpose

of safeguards is to deter or reveal national intentions to obtain nuclear weapons

capability by providing the time required for abortive international actions to be

taken.

The U. S. Department of Energy has instituted a variety of programs to establish

nonproliferation criteria and develop fuel cycle technologies and safeguards systems

consistent with these criteria. This year the Department of Energy Nonproliferation

i

.:

lt:(

1 fc:

Li!9

I:

i

s ii

78
ternative Systems Assessment Program will have completed a review of literally scores alternatives to the existing LWR uranium-plutonium fuel cycle. At that time those

chnically feasible and economically practical fuel cycles which may offer superior

nproliferation features will be identified and considered for possible development

ward commercialization objectives.

In a larger context, some forty nations are currently participating in the

ternational Fuel Cycle Evaluation Committee which has a similar two year charter

examine alternatives to plutonium reprocessing and recycle and establish relative

nproliferation features. U. S. support for this activity is provided by the DOE

SAP program and direct participat ion in the various INFCE working groups , including

put of the U. S. technical information needed to allow in-depth consideration of the

ternatives

.

In support of the U. S. nonproliferation efforts, the Department of Energy

feguards and Security program is currently developing safeguards concepts, techlogies, and integrated systems that can be applied to any of the future fuel cycles.

addition, methodologies have been developed and are being utilized in selected

ternatives to quantitatively determine the inherent detection timeliness and sensivity features of those fuel cycles. These results will be compared with our studies

the LWR fuel cycle and the conclusions submitted to the NASAP and INFCE programs.

SAFEGUARDS AND NONPROLIFERAHON

Regardless of the detailed form of the future domestic commercial and international

.el cycles, we believe that certain safeguards considerations are inherent to proferation resistance. The most critical area of the current fuel cycles is the postactor segment including reprocessing, conversion and fuel ref abrication. Whether the

.ture fuel cycle is based on coprecipitated plutonium, thorium or spiked recycle

.el materials, safeguarding the back end of the fuel cycle will still place its

eatest demand on the technological capabilities available to be used in that area.

The selection of future fuel cycles on the basis of nonproliferation criteria

.11 depend, to an extent, on the passive safeguards features inherent to those fuel

cles and the active measures that can be additionally made available to achieve

equate levels of safeguards effectiveness. Passive safeguards features that differ

om one fuel cycle to another include levels of biologically hazardous constitutents

,

temical and physical form, and the complexity of processing required to convert

lei cycle material to weapons usable form. The Civex fuel cycle and its variants,

•r example, possess the passive safeguards feature of biologically hazardous, relatively

lattractive fuel cycle materials, since recycled fissile material is never completely

•.parated from fission-product contaminants.

Although minimizing the attractiveness of nuclear material in future fuel cycles is

desirable feature, the choice of proliferation-resistant fuel cycles for domestic and

vternational energy production will undoubtedly be tempered by environmental, economic,

id political considerations. Fuel cycles with potentially high environmental risks,

ich as high radioactivity levels and possible criticality accidents, are not likely

> be acceptable as nuclear energy options. As an additional nonproliferation tool

i believe that active safeguards measures such as continuous material monitoring and

iasurement, surveillance, and automatic process controls, can be added to the passive

iatures

.

We believe that the technologically and economically feasible, and environmentally

:ceptable fuel cycles of the future will require active safeguards systems to supplest the inherent, passive safeguards features. Because of this belief, our developantal program is structured to culminate in integrated safeguards systems capable of

ial-time control and continuous process monitoring.

79
To satisfy the demands of proliferation resistance, fuel cycle safeguards must

be based on continuous monitoring of nuclear material from the reactor through reproces;

ing, ref abrication and back to the reactor. Only by continuous monitoring can un- controlled build-up or diversion of weapons material be prevented. In our view, the

principal safeguards element for continuously monitoring irradiated and refabricated

fuel assemblies is containment, surveillance, and inspector verification of shipment

and reception. Since these fuel assemblies are either highly radioactive or denatured

to low enriched levels, the technology, systems, and institutional arrangements for the! safeguards are currently available. Whether the fuel cycle is uranium/plutonium,

uranium/ thorium, or any of the alternatives, this portion of the fuel cycle can be

safeguarded by today's methods.

Continuous monitoring of nuclear material during reprocessing, chemical conversion

and fuel refabrication is substantially more difficult. During these processing j operations, the form and concentration of nuclear material becomes relatively more

attractive as a diversion target. In the nonproliferation context, the principal

safeguards elements are the active measures: continuous process monitoring, surveillan;

and real-time accounting. As the nuclear material form and concentration become more

attractive, increasingly stringent safeguards are required, that is, better measurements, tighter administrative control, and more timely detection and detailed process i monitoring. The ability to monitor the movement of nuclear material by active safe- ' guards measures during these processing operations, may well decide the fate of many J of the currently proposed alternative fuel cycles.

SAFEGUARDS AND SECURITY PROGRAM

Our safeguards concepts and systems are based on a "mini-max" principle:

minimizing the opportunity for diversion and maximizing the probability of timely

detection to allow for the necessary communication and control actions. The ability

to continuously monitor and control nuclear material during processing in accordance

with this principle is the final goal of our research and development program. It

is believed that all fuel cycles must inherently be affected, if not governed, by

the adoption of these active safeguards measures.

Of key importance to the nuclear energy option is the cost-effectiveness of futur

safeguards systems. In all of our efforts, effectiveness must be measured against

design, implementation and operational costs. There can be no doubt, however, that

a modern, automated safeguards system substantially increases a plant operator's

ability to monitor and optimize the efficiency of his processing operation. Cost

effectiveness is the corollary to our "mini-max" principle of safeguards system design ,

The SS plan for developing integrated safeguards systems consistent with national ; proliferation objectives is composed of three phases:

• Generic conceptual design;

• Technology and subsystem development testing and evaluation; and

• Integrated safeguards system design testing and evaluation.

The generic conceptual design for safeguarding a particular facility or alternati

fuel cycle represents a preliminary scoping, definition and analysis of the safeguards

problems. The generic conceptual design establishes basic engineering design criteria

and considerations for the incorporation of effective safeguards, and outlines preliminary material control and accounting and physical protection systems. In additior

these studies identify available and reasonably projected safeguards technology and

those subsystems requiring further research and development. In direct support of

U. S. nonproliferation objectives, we are currently adapting our generic safeguards

concepts to the alternative fuel cycles under DOE technology development programs.

This effort will address all aspects of the fuel cycles including reprocessing,

j6i

111 ire

80
fabrication, multi-national fuel centers, denatured and biologically hazardous

lear material forms and concepts for international inspection. Generic studies

two alternatives, coprecipitation and uranium- thorium will be completed this year.

The technology and subsystem development phase of our program includes the

velopment of nuclear material measurement methods and instrumentation, process

nitoring and control modules and subsystems, data management and analysis methodology , and physical protection, surveillance, containment subsystems. Individual

bsystem development projects, such as specific measurement instrumentation, culminate

rigorous, well-defined test and evaluation programs in operating nuclear facilities,

the development of nuclear material measurement methods, we have reached the stage

ere reliable, prototype instrumentation can be engineered and installed in the

ocessing lines of today's nuclear facilities. Examples of recently developed protope instrumentation include the absorption-edge densitometer and californium-252

uffler to be discussed later during this conference. These technological developnts are the individual building blocks for the improvements at existing facilities

-d for the integrated safeguards systems of the future.

Recently, the joint project between the U. S., and Japan, with participation

I France and the IAEA has been agreed upon to test advanced safeguards instrumentaon and methods at the Tokai-Kura reprocessing facility in Japan. Thirteen tasks

.ve been established, inclusively covering surveillance and containment at the

.orage pond to measurement and monitoring systems at the product end of the facility.

though complete evaluation results will not be available until the end of calendar

:ar 1979, this joint effort is expected to produce results that will be available

>r IKFCE considerations.

The third phase of safeguards system development in our program is the integrated

'stem design testing and evaluation. Our goal for integrated safeguards system

;velopment is to achieve real-time accounting supplemented by continuous process

mitoring in accordance with the "mini-max" principle and its corollary, cost-

: f ectiveness . In this phase, we assemble the basic components developed under Phase 2 lto a cost-effective system and test the entire system under operating conditions

: a nuclear facility. Based on detailed engineering designs, safeguards systems

)deling and simulation are used to establish safeguards system effectiveness, relative

-)st-effectiveness of alternative systems and in the case of alternative fuel cycles,

stablish a quantitative comparison of effectiveness, cost, and inspectability with

cisting LWP fuel cycles.

These efforts have culminated in the implementation of advanced material control

]id accounting systems at several Department of Energy facilities. Real-time material

nntrol and accounting is currently beginning test and evaluation at the plutonium

•rocessing and recovery facility, TA-55 at Los Alamos, and the Y-12 facility at Oak

idge. Similar systems based on this concept are in the planning stages for other

acilities. We are also beginning to test an advanced process monitoring system at

ie Idaho Chemical Processing Plant that will enable continuous monitoring of bulk

aterial flow throughout a reprocessing facility.

CONCLUSIONS

The fuel cycle safeguards concepts, technology, and advanced, integrated systems

nder our development programs are emerging as factors that will shape the future of

he nuclear energy option.

81
The deferral of domestic reprocessing and recycle and the associated U. S. policy

initiatives in the international nuclear community can afford us time to make available

practical active safeguards for assuring the world that nuclear materials are being

used only for peaceful purposes. This is the goal underlying today's conference.

i later:

,; is;

!; citing

:

i 1-5

J iily i

.i lie i

i •

f

!

i In tl

£ able us i

' >ichi

i divert

< of tt

? tylii

11 These

jitkat

: ia th

!' lines
Decision Analysis for Dynamic Accounting of Nuclear Material

by

J. P. SHIPLEY

Los Alamos Scientific Laboratory, Los Alamos, New Mexico

ABSTRACT

Effective materials accounting for special nuclear material in

modern fuel cycle facilities will depend heavily on sophisticated dataanalysis techniques. Decision analysis , which combines elements of

estimation theory, decision theory, and systems analysis, is a framework well suited to the development and application of these techniques.

Augmented by pattern-recognition tools such as the alarm-sequence chart,

decision analysis can be used to reduce errors caused by subjective

data evaluation and to condense large collections of data to a smaller

set of more descriptive statistics. Application to data from a model

plutonium nitrate-to-oxide conversion process illustrates the concepts.

KEYWORDS: Nuclear safeguards, materials accounting, decision analysis,

CUSUM, Kalman filter, alarm-sequence chart, plutonium

nitrate-to-oxide conversion

INTRODUCTION

Materials accounting for safeguarding special nuclear material (SNM) has two important

its: (1) the collection of materials accounting data, and (2) the analysis of materials

nting data. The collection function is a broad, highly developed subject (e.g., see

| 1-5 and the references therein) that we will not pursue here; in this paper we are

rily concerned with the analysis of materials accounting data.

Ihe data collection function is usually structured to facilitate performance of the

sis function, commonly by providing sufficient measurements of SNM within a facility

low the drawing of material balances around selected portions of the facility on a nable time scale. The data, which are always corrupted by measurement errors, often

r as time sequences of material balances, one sequence from each part of the facility

hich material balances are drawn.

Therefore, the data-analysis function must operate on imperfect data that become availsequentially in time. Its primary goals are (1) detection of the event (s) that SNM has

diverted, (2) estimation of the amounts diverted, and (3) determination of the signifiof the estimates. Furthermore, data analysis must search for evidence of diversion

may have occurred in any of several patterns.

These goals are ideal for statistical treatment using sequential, probabilistic tefhs that have been developed for applications to communications and control systems.

ion analysls ^'^ is a framework of such tools, and combines techniques from estimation

y and hypothesis testing, or decision theory, with systems analysis for treating com- dynamic problems. The decision-analysis framework is general enough to allow a wide

in the level of sophistication in examining SNM accounting data, while providing

lines for the development and application of a variety of powerful methods.

83
THE DECISION PROBLEM

Let us suppose that over some time period we have acquired materials accounting data- consisting of a set of in-process inventory measurements at discrete times and a set of material transfer measurements between those times. Let I(k) be the k

t '1 inventory measuri ment, and let T(k) be the measurement of the net transfers that occurred between the I(k) and I(k+1) inventory measurements. If the measurements were exact and there had been no diversion of SNM, then the continuity equation would be satisfied:

I(k+1) = I(k) + T(k) .

However, we never have perfect measurements of bulk material, and SNM may or may not have been diverted, so that (1) should be rewritten as

U

is o

acc

Los

nt

I(k+1) = I(k) + T(k) - M(k+1) ,

where M(k+1) is the imbalance in the continuity equation (1) at time k+1 caused by measui

ment errors and diversion. We call M(k) the material balance value at time k, or the k"

material balance for short. Clearly, M(k) is a random variable, and the sequence

{M(k), k = 2,3,...} is a stochastic process, about which we can make probabilistic stated

ments if something is known of the nature of the measurement errors.

For convenience, denote the set of inventory measurements {l(k), k = 1,2,...,N}, th|| set of net transfer measurements {T(k), k = 1, 2 , . . . ,N-l} , and the statistical information

the measurement errors by the quantity Z(N). Then Z(N) contains all information availab 1 for the N materials accounting intervals. Thus, the decision problem is to determine, b! on observation of Z(N), whether the set {M(k+1) , k = l,2,...,N-l} (M^ for simplicity) co tains diversion.

The Likelihood Ratio

For any particular Z(N) that is observed, M^ may or may not have contained diversic

Define the hypotheses

3 61

If

H: M^ does not contain diversion,

K: M^ does contain diversion.

Then the actual probability density function that governs Z(N) is determined by which oi H, K is true; i.e.

,

Z(N) ^ p[Z(N)|H] for H true,

Z(N) % p[Z(N)|K] for K true,

where ^ means "has the density function." These two conditional density functions are called likelihood functions for the hypotheses H and K. The values of the likelihood fi tions for a particular Z(N) are relative measures of the likelihood that Z(N) is govern

one or the other density function, or in other words, that H is true or K is true.

The usual statistical test consists of forming the likelihood ratio , ^

^ L, and com

ing it to a threshold:

a ri/-a\\vi i < T > accept H,

if L[z ( N)] a

i

*

p [z(n)|h] ;

y > T> accept K>

where T is the threshold to be determined below. Roughly speaking, if Z(N) is "enough"

likely to have occurred as a result of H being true than of K being true, then decide t is true; otherwise, decide that K is true.

84
Conversion to Sufficient Statistics

5 likelihood functions are difficult to work with because they are joint density

{as of many variables, in general. Under certain circumstances, which usually hold

: accounting, it is possible to condense the quantity Z(N) to a single number S(N)

I loss of information. The number S(N) is called a sufficient statistic 7 and is

fsnt to knowledge of Z(N). If such a S(N) can be found, and if its density function

calculated, then the likelihood ratio test (3) can be replaced by

If L[S(N)] = p[S(N) K]

p[S(N) H]

< T\ accept H,

> T', accept K.

(4)

a density (i.e., likelihood) functions are univariate and, therefore, much more

Le mathematically. The hidden problem is to find a sufficient statistic that is

;antly indicative of whether H or K is true.

Sequential Formulation

we have seen, the likelihood ratio test, (3) or (4), for a fixed number N of points

s of comparing the likelihood ratio to a single threshold. However, in practical

pns we seldom will know when the test should start or end. Therefore, we will want

~.i the test at all possible starting points and let the test itself determine when it

"oe terminated. This procedure also has the provident property of requiring fewer

", on the average, than a fixed-sample-size test having the same characteristics. 8

It the sequential J ikelihood ratio test, also called the sequential probability ratio

SPRT, 8 there are three possible results at each decision time, rather than two:

< Tq, accept H, If L[S(k)] | > accept K, (5)

otherwise, take another observation,

SPRT is repeated for all possible starting points. The thresholds Tq and T^ can be

rom the Neyman-Pearson criterion or by minimizing the Bayes risk,? but that may

some information that is unavailable. The following approximation, devised by Wald,

seful thresholds that can be shown to be conservative.

: Pf4 and Pp be the desired (given) miss and false-alarm probabilities, respectively,

SPRT. Then the thresholds are 7 * 8

T =

0 1 - p f

(6)

T

1 " PM

1 P

F

.aability of detecting diversion, related to l-P^, is called the power or size of the

v is called the significance or level of the test.

if SOME SUFFICIENT STATISTICS

r any decision problem, there is a large number of sufficient statistics that may be

ted, but some are more useful than others because of computational reasons, closer

. nship to physically meaningful quantities, or better discriminatory powers between

hypotheses. Following are several statistics that have been effective in various

tions

.

85
The CUSUM Statistic

The CUSUM"'" '"^ (cumulative summation) of material balances is just the sum of the! material balances over the time period of interest.

CUSUM(k+l) = M(2) + M(3) + ... + M(k+1) ,

where the M(i) are found from (2)

M(i+1) = - I(i+1) + I(i) + T(i) , i = 1,2,. ..,k .

The CUSUM can also be written as

k

CUSUM(k+l) = - I(k+1) + 1(1) + Z T(i) , i=l

which emphasizes that the material balances are negatively correlated through the comnio inventory measurement. For uncorrelated transfer measurements, the CUSUM variance is

k

VC(k+l) = VI(k+l) + VI (1) + E VT(i) , i=l

where VI(*) and VT(*) are the inventory and transfer measurement error variances, resp« tively In recursive form suitable for a SPRT, the cusum and its variance can be writt: as

CUSUM(k+l) = CUSUM(k) - I(k+1) + I(k) + T(k)

VC(k+l) = VC(k) + VI(k+l) - VI (k) + VT(k) .

The corresponding SPRT can be shown to reduce to

f CUSUM (k+l)

/VC(k+l)

, £ - /2|ln T | , accept H,

•ice > + /2]ln T | , accept K,

( otherwise, take another observation.

The CUSUM statistic is interesting because it is an estimate of the total amount of mi

material during the period. However, the CUSUM is not a minimum-variance statistic un the variances of the material balance measurements are all equal, and unless the knowl' of how the material balances combine inventory and transfer measurements is unimportaii

last condition would hold if the inventory were small or well measured compared to the fers.

The Two-State Kalman Filter Statistic

2 17-20

The two-state Kalman filter statistic ' estimates the average amount of mis; material per balance. It uses all available information from the continuity equation

and from the statistics of the measurement errors. The two-state Kalman filter statit can be shown to be optimal in the sense that it is the minimum-variance, unbiased, lir estimate whenever the measurement error probability densities are symmetric about the} means.

The two-state Kalman filter yields estimates of both the inventory and the mater: balance at each time. In recursive form, the equations are^l>22

ttffi

liiij

86
I(k+1) = I(k+l|k) + K [I(k+1) - I (k+1 | k) ] ,

M(k+1) = M(k) + K

2 [I (k+1) - I (k+1 | k)] ,

(10)

I(k+l|k) = I(k) + T(k) - M(k) ,

1) and M(k+1) are the inventory and material balance estimates, respectively, at based on all information through time k+1. The filter gains, and K^, are given

= VIE (k+1)

1 VI (k+1)

_ VMIE(k+l)

2 VI (k+1)

(ID

E(k+1) and VMIE(k+l) are respectively the inventory estimate error variance and the

ice between the inventory and material balance estimate errors. They are given

; ely by

VIE (k+1) = VIE (k+1 | k) VI (k+1)

VIE (k+1 Ik) + VI (k+1)

= VMlE(k+ljk) VI (k+1) raiMK+JJ

VIE (k+1 k) + VI (k+1) '

(12)

VIE (k+1 | k) = VIE(k) - 2VMIE(k) + VME(k) + VI (k)

VMIE(k+l|k) = VMIE(k) - VME(k) . (13)

ance of the error of the material balance estimate at time k+1, VME(k+l), is

VME(k+l) = VME(k) - VMIE (k+l|k)

VIE (k+1 [k) + VI (k+1)

'

rences 2 and 17-24 for more detail,

resulting SPRT is similar to that for the CUSUM, and reduces to

(14)

If M(k+1)

/VME(k+l)

<_ - /2 | In T

Q

| , accept H,

> + /2|ln T \ , accept K,

otherwise, take another observation.

(15)

Other Sufficient Statistics

sufficient statistics such as those just discussed are called parametric because

end upon knowledge of the statistics of the measurement errors. They also happen to

t when the measurement errors are Gaussian, a quite common but by no means all-

: 'e situation. If the measurement error statistics are unknown or non-Gaussian, then

etric_25 sufficient statistics may give better test results. In addition, nonparaests can provide independent support for the results of parametric tests even though

etric tests are generally less powerful than parametric ones under conditions for

e latter are designed.

87
The two most common nonparame trie tests are the sign test and the Wilcoxon rank sun test. The sufficient statistic for the sign test is the total number of positive materia balances. That for the Wilcoxon test is the sum of the ranks of the material balances.

The rank of a material balance is its relative position index under a reordering of the material balances according to magnitude. Other, possibly more effective, nonparametric

tests are being investigated. Further discussion of nonparametric tests is beyond the scope of this paper.

TEST APPLICATION

Procedure

As discussed above, we seldom will know beforehand when diversion has started or hoi I long it will last. Therefore, the decision tests must examine all possible, contiguous

sequences of the available materials accounting data. That is, if at some time we have material balances, then there are N starting points for N possible sequences, all ending 1 the Nth, or current, material balance, and the sequence lengths range from N to 1. Beca, of the sequential application of the tests, sequences with ending points less than N hav

1 already been tested; those with ending points greater than N will be done if the tests d'1" not terminate before then.

Another procedure that helps in interpreting the results of tests is to do the test at several significance levels, or false-alarm probabilities. This is so because, in pi tice, the test thresholds are never exactly met; thus, the true significance of the dat; obscured. Several thresholds corresponding to different false-alarm probabilities give

least a rough idea of the actual probability of a false alarm.

Displaying the Results

Of course, one of the results of most interest is the sufficient statistic. Coramo practice is to plot the sufficient statistic, with symmetric error bars of length twice

square root of its variance, vs the material balance number. The initial material bala and the total number of material balances are chosen arbitrarily, perhaps to correspond

the shift or campaign structure of the process. For example, if balances are drawn hoi and a day consists of three shifts, then the initial material balance might be chosen i the first of the day, and the total number of material balances might be 24, covering t shifts. Remember, however, that this choice is for display purposes only; the actual ; ing procedure selects all possible initial points and sequence lengths, and any suffic

statistic may be displayed as seems appropriate.

The other important results are the outcomes of the tests, performed at the sever; significance levels. A new tool, called the alarm-sequence chart, 1~3,12 ^as v, een deve. to display these results in compact and readable form. To generate the alarm-sequence

each sequence causing an alarm is assigned (1) a descriptor that classifies the alarm

ing to its false-alarm probability, and (2) a pair of integers (r^,^) that are, respe

tively, the indexes of .the initial and final material balance numbers in the sequence,

alarm-sequence chart is a point plot of r-^ vs ^ for each sequence that caused an alar

with the significance range of each point indicated by the plotting symbol. One possi

correspondence of plotting symbol to significance is given in Table I, The symbol T m

* It is also possible to define (r-^,^) as the sequence length and the final material

ance number of the sequence. The two definitions are equivalent.

'irnac

oat f

88 iti
TABLE I

ALARM CLASSIFICATION FOR THE ALARM-SEQUENCE CHART

Class if ication

(Plotting Symbol) False-Alarm Probability

A 10-2 to 5 x 10-3

B 5 x 10-3 t o 10-3

C lO-3 to 5 x io-4

D 10"J to 5 x 10*^

5 x 10~4 to 10~4 E 10-4 to 10-5 F < 10~5 T ^ 0.5

, jences of such low significance that it would be fruitless to examine extensions of them;

letter T indicates their termination points. It is always true that r^ £ r£ so that all

ools lie to the right of a 45° line through the origin. Examples of these charts are

[*n in the section on results.

AN EXAMPLE

The Process

To illustrate the application of decision analysis, we present results from a study of

trials accounting in a plutonium nitrate-to-oxide conversion facility. 3 The reference

;ess is based on plutonium (Ill)-oxalate precipitation; a simplified block diagram is

<m in Fig. 1. Nominal capacity is 116 kg of plutonium per day, processed in 2-kg

;hes. Some of the most important design parameters for the main process stream are given

Table II.

TABLE II

CONVERSION PROCESS DESIGN PARAMETERS

Function

Volume or Weight

leceipt tank feed

/alence adjust feed

Precipitator feed

?u oxalate boat to furnace

,?u oxide to accountability

3 u product to storage

7 iltrate

precipitator flush

"urnace sweeping

3oat flush

Dump station sweep

Per Batch Concentration Frequency

200.0 L 30.0 g/L 1/0.41 h

66.67 L 30.0 g/L 1/0.41 h

75.44 L 26.5 g/L 1/0.41 h

4.65 kg 0.422 kg/kg 1/0.41 h

2.21 kg 0.882 kg/kg 1/0.41 h

2.18 kg 0.882 kg/kg 1/0.41 h

154.9 L 66.4 mg/L 1/0.41 h

109.2 L 4.6 g/L 3 /day

0.85 kg 0.882 kg/kg 1/week

34.4 L 2.9 g/L 10/day

0.85 kg 0.882 kg/kg 2 /day

The Materials Accounting System

Many different ways of drawing material balances for the conversion process can be

Lned. Based on the conversion study, 3 one strategy that works very well is to consider

main process stream from the receipt tank to the product dump and assay station as one

: process. Thus, the transfers consist of feed into the receipt tank, product out of the

) and assay station, and recycle solids and liquids. All these transfers must be meas1, and we must obtain an estimate of the in-process inventory. Table III gives the

lired measurements and some possible measurement methods and associated uncertainties.

89
TABLE III

MATERIALS ACCOUNTING MEASUREMENTS FOR THE CONVERSION PROCESS

Measurement Point

Receipt tank

Wet boat

(precipitator output)

Precipitator holdup

Filtrate

Precipitator flush

Boat flush

Furnace sweep

Dump station sweep

Product cans

Measurement Type

Volume

Concentration (by L-edge

densitometry)

Mass (by neutron well counter)

Mass (by He-3 neutron counter)

Volume

Concentration (by alpha monitor)

Volume

Concentration (by L-edge

densitometry)

Volume

Concentration (by x-ray

fluorescence)

Mass (by neutron well counter)

Mass (by neutron well counter)

Mass (by neutron well counter,

calorimeter, or gamma spectrometer

Instrument

Precision

(%)

0.2

1

0.2

10

0.2

1

0.2

1

2

2

1

Calibration Error

(%)

0.1

0.3

0.1

20.1

0.3

0.1

0.3

0.5

0.5

0.5

See References 1-3 for detailed discussions of measurement techniques.

Results

The techniques of decision analysis described earlier have been used to evaluate the

diversion sensitivity of this materials accounting strategy and others. Part of the evalua

tion consists of examining test results, in the form of estimate (sufficient statistic) and alarm-sequence charts, for various time intervals. Examples of typical one-day charts for

the CUSUM and two-state Kalman filter are shown in Figs. 2 and 4; the corresponding alarmsequence charts are given in Figs. 3 and 5.-^ In the course of evaluation, many such sets c charts are examined so that the random effects of measurement errors and normal process

variability can be assessed; that is, we perform a Monte Carlo study to estimate the sensitivity to diversion. In applying decision analysis to data from a facility operating undei

actual conditions, only one set of data will be available for making decisions, rather thar the multiple data streams generated from a simulation. The decision-maker will have to extrapolate from historical information and from careful process and measurement analysis

ascertain his true diversion sensitivity.

The results of the evaluation are given in Table IV. By comparison, current regulatii

require that the material balance uncertainty be less than 0.5% (2a) of throughput for eacl two-month accounting period, which corresponds to 33 kg of plutonium for this process. Sun large improvement in diversion sensitivity is possible through the combination of timely

measurements with the powerful statistical methods of decision analysis.

90
TABLE IV

DIVERSION SENSITIVITY FOR THE CONVERSION PROCESS

Detection Time

Average Diversion Per Batch

(kg Pu)

Total at Time of Detection

(kg Pu)

1 batch (1.35 h)

1 day

1 week

1 month

0.13

0.03

0.01

0.005

0.13

0.63

1.24

2.65

ACKNOWLEDGMENT

The author greatly appreciates the assistance and contributions of his colleagues in

Safeguards Systems Group (Q-4) of the Los Alamos Scientific Laboratory.

J. P. Shipley, D. D. Cobb, R. J. Dietz, M. L. Evans, E. P. Schelonka, D. B. Smith, and

R. B. Walton, "Coordinated Safeguards for Materials Management in a Mixed-Oxide Fuel

Facility," Los Alamos Scientific Laboratory report LA-6536 (February 1977).

E. A. Hakkila, D. D. Cobb, H. A. Dayem, R. J. Dietz, E. A. Kern, E. P. Schelonka,

J. P. Shipley, D. B. Smith, R. H. Augustson, and J. W. Barnes, "Coordinated Safeguards

for Materials Management in a Fuel Reprocessing Plant," Los Alamos Scientific Laboratory report LA-6881 (September 1977).

H. A. Dayem, D. D. Cobb, R. J. Dietz, E. A. Hakkila, E. A. Kern, J. P. Shipley, D. B. Smith, and D. F. Bowersox, "Coordinated Safeguards for Materials Management in a Nitrate-to-Oxide Conversion Facility," Los Alamos Scientific Laboratory report LA-7011

(to be published)

G. R. Keepin and W. J. Maraman, "Nondestructive Assay Technology and In-Plant Dynamic

Materials Control—DYMAC," in Safeguarding Nuclear Materials , Proc. Symp. , Vienna,

Oct. 20-24, 1975 (International Atomic Energy Agency, Vienna, 1976), Paper IAEA-SM201/32, Vol. 1, pp. 304-320.

R. H. Augustson, "Development of In-Plant Real-Time Materials Control: The DYMAC Program," Proc. 17th Annual Meeting of the Institute of Nuclear Materials Management,

Seattle, Washington, June 22-24, 1976.

W. L. Root, "An Introduction to the Theory of the Detection of Signals in Noise," Proc.

IEEE, Special Issue on Detection Theory and Applications .58, No. 5, 610-623 (1970).

A. P. Sage and J. L. Melsa, Estimation Theory with Applications to Communications and

Control (McGraw-Hill, 1971).

A. Wald, Sequential Analysis (John Wiley and Sons, Inc., 1947).

R. Deutsch, Estimation Theory (Prentice-Hall, 1965).

R. A. Howard, "Decision Analysis: Perspectives on Inference, Decision, and Experimental

tion," Proc. IEEE, Special Issue on Detection Theory and Applications _58, No. 5, 632-

REFERENCES

643 (1970).

Ref. 2, Vol. II, Appendix E.

91.
12. D. D.Cobb, D. B. Smith, and J. P. Shipley, "Cumulative Sum Charts in Safeguarding

Special Nuclear Materials," submitted to Technometrics (December 1976).

13. A. J. Duncan, Quality Control and Industrial Statistics (R. D. Irwin, Inc., 1965),

14. E. S. Page, "Cumulative Sum Charts," Technometrics _3, No. 1, 1-9 CFebruary 1961).

15. W. D. Evans, "When and How to Use Cu-Sum Charts," Technometrics 5_, No. 1, 1-22

(February 1963)

16. J. M. Lucas, "A Modified V-Mask Control System," Technometrics L5, No. 5, 833-847

(November 1973)

17. D. H. Pike, G. W. Morrison, and C. W. Holland, "Linear Filtering Applied to Safeguards

of Nuclear Material," Trans. Amer. Nucl. Soc. 22^, 143-144 (1975).

18. D. H. Pike, G. W. Morrison, and C. W. Holland, "A Comparison of Several Kalman Filter

Models for Establishing MUF," Trans. Amer. Nucl. Soc. _23, 267-268 (1976).

19. D. H. Pike and G. W. Morrison, "A New Approach to Safeguards Accounting," Oak Ridge

National Laboratory report ORNL/CSD/TM-25 (March 1977).

20. D. H. Pike and G. W. Morrison, "A New Approach to Safeguards Accounting," Nucl. Mater.

Manage. VI(III) , 641-658 (1977).

21. R. E. Kalman, "A New Approach to Linear Filtering and Prediction Problems," Trans,

ASME J. Basic Eng. 82D, 34-45 (March 1960).

22. R. E. Kalman and R. S. Bucy, "New Results in Linear Filtering and Prediction Theory,"

Trans. ASME J. Basic Eng. 83D, 95-108 (March 1961).

23. J. S. Meditch, Stochastic Optimal Linear Estimation and Control (McGraw-Hill f 1969),

24. A. H. Jazwinski, Stochastic Processes and Filtering Theory (Academic Press, 1970),

25. J. B. Thomas, "Nonparametric Detection," Proc. IEEE, Special Issue on Detection Theory

and Applications 5j3, No. 5, 623-632 (May 1970).

NOTICE

"This report was prepared as an account of work

sponsored by the United States Government.

Neither the United States nor the United States Department of Energy, nor any of their employees, nor any of their contractors, subcontractors, or

their employees, makes any warranty, express or

implied, or assumes any legal liability or responsibility for the accuracy, completeness or

usefulness of any information, apparatus, product

or process disclosed, or represents that its use would not infringe privately-owned rights."

92
RECEIPT

TANK

PRODUCT

STORAGE

Fig. 1. Simplified block diagram of the plutonium nitrate-tooxide conversion process.

9 3
5 10 15 20

Materual Balance Number

5 10 15 Material Balance Number

Fig. 2. CUSUM charts for one day with no diversion (upper),

with diversion (lower)

9 A
co

Q_

o

-J

._>c

5 10 15 Penal PoLnt

20

-<-> c

. _>o

Q_

O

._)

5 10 15 PLnaL PoLnt

Fig. 3. Alarm-sequence charts for CUSUM: with no diversion

(upper), with diversion (lower).

95
3

QuCD

CO

o

. _>L

CD

cnc

._>

COCO

. _)

0-2

0- 1 --

0-0

-0- 1 --

-0-2

5 10 15 20

Material Balance Number

3

cn

o

e

w

LJ

COc

. _>

CO

CO

. -J

0-2

0- 1 --

0-0

-0- 1

-0-2

5 10 15 20

Material Balance Number

Fig. 4. Kalraan filter estimates of average missing material

for one day with no diversion (upper) , with diversion

(lower)

96
20

10 15 20

Final Point

20

C

._>o

Q_

. _>C

15 --

10 --

5 --

n a B fl B R

B B

D D

5 10 15 Final Point

20

Fig. 5. Alarm-sequence charts for Kalman filter estimates

with no diversion (upper) , with diversion (lower)

97
Analytical Chemistry Measurements Quality

Control Program Using Computer Applications

by

J. P. CLARK and G. A. HUFF

Allied-General Nuclear Services, Barnwell, South Carolina 29812

ABSTRACT

Lar;

COO'

its

- Pre;

ty c

An Analytical Chemistry Measurements Quality Control Program

assures the reliability of analytical measurements performed at the

Barnwell Nuclear Fuel Plant. The program includes training, methods

quality control, replicate samples and measurements, mass measurements, interlaboratory sample exchanges, and standards preparation.

This program has been designed to meet the requirements of 10CFR70.57. ( ieta Portions of the program have been automated by using a PDP 11/35

computer system to provide features which are not readily available

in manual systems. These include such items as realtime measurement control, computer calculated bias and precision estimates,

various surveillance applications, and evaluation of measurement

system variables.

The efficiency of the computer system has been demonstrated in

gathering and assimilating the results of over 1100 quality control

samples during a recent cold chemical checkout campaign. These data were used to determine equations for predicting measurements reliability estimates; to evaluate measurement performance of the analysts,

equipment, and measurement period; and to provide directions for

chemistry methods modifications and additional training requirements.

A procedure of replicate sampling and measuring provides random

error estimates. The analytical chemistry measurement quality

control activities during the campaign represented about 10% of

the total analytical chemistry effort.

KEYWORDS: Assurance; measurement quality control; realtime quality

control; reliability; replicates; standards

INTRODUCTION

This paper describes the Analytical Chemistry Measurements Quality Control Progr

used at the Barnwell Nuclear Fuel Plant located in Barnwell, South Carolina. This pi was built to recover uranium and plutonium from spent light water reactor fuel. At

capacity operation, the plant will daily process five metric tons of uranium and have over 200 samples taken, requiring 800 analytical determinations per day. Analytical

vices Department personnel are responsible for these determinations, which are made f controlling plant processes, calibrating in-line monitors, and accounting for nucleat

materials. The quality of these measurements is of utmost importance. Therefore, a surement quality control program has been designed to determine the reliability of tr analytical measurements. The program helps identify faulty methods, equipment, anal}

techniques, reagents, and systems which require remedial attention.

A "Measurement Control Program for Special Nuclear Materials Control and Account

(10CFR70.57) is required by the Nuclear Regulatory Commission. In addition to nuclef

material measurements, the Barnwell program includes all routine measurements. The i program includes the following segments: « "

i 1. A procedure for training and testing the technicians.

2. A procedure for analytical chemistry methods quality control.

3. A procedure for replicate samples and measurements.

98

sta:

ze n

tie

IS (

lyzi

To ac

ltt

ates

ices,

vers

of
A procedure for mass measurements and balance quality control.

5. Participation in an interlaboratory sample exchange program.

6. The preparation of reagents and standards.

Large portions of segment two of the program have been automated by using the laboracomputer system. The use of the computer system will be discussed in detail. Other

nts of the program will be addressed briefly.

THE ANALYTICAL CHEMISTRY METHODS QUALITY CONTROL PROCEDURE

Presently some 25 analytical chemistry methods are included in this segment of our

ty control program. Each method has a stated range for which the method is .appliStandards are prepared to cover the range of each method. Usually several pairs

andards are made up a couple of standard deviations apart to cover the range. These

sed in the quality control program to test the various levels over the range. In most

these standards are packaged in glass ampuls and flame sealed. Currently, the StanLaboratory has over 50,000 standards packaged in flame sealed ampuls for use in the

ty control and training activities. The reference values for these standards are ietary and known only to the Standards Laboratory personnel.

Technicians are trained to use the analytical chemistry methods by analyzing a known

standard until they become proficient with the method. Then they are required to

ze nine unknown standards. If statistical evaluation of the results proves satisfacthe technician is qualified to use the method for the analyses of plant samples.

Routinely, each shift an analytical chemistry method is used, the analyst is required

alyze a quality control standard. If this measurement is within the control limits

he method, he is permitted to use the method for plant sample measurements. However,

is out of control, the measurement system must be corrected before plant samples can

alyzed.

To achieve this requirement, the laboratory computer system has been programmed to ol the quality of our analytical methods measurements. The computer monitors each

atory method and will lock out the calculation program for a method if a standard

ates the method is out of control. A description of this system is given.

Laboratory Computer System Description

The laboratory computer system as shown in Figure 1, consists of a PDP 11/35 CPU with

of memory, multiple disk drives, dual magnetic tape transports, and various input/

it terminals (DECwriters and CRT terminals). One-fourth of the disk storage provides

5000 files for the quality control standards activities. Periodically, data from

r zed standards are transferred to magnetic tape for long term storage. Terminals are

:ed at strategic points in the laboratories, process control rooms, and the administrabuilding. Initially, only two laboratory instruments were interfaced directly to the

:m, a mass spectrometer and a multichannel analyzer. We are currently interfacing

ices, densimeters, f luorophotometers, automatic titrators, and other laboratory instruThe system is built around the RSTS/E (resource sharing-time sharing) operating sys-

, version 6B, and utilizes a BASIC-PLUS language processor. All applications programs

:he individual analytical methods are written in a modular format using this powerful

Lon of BASIC. This system has been operating successfully for two years. During this

id we have experienced less than 1% unscheduled downtime.

The plant recently completed a 10-week cold run during which some 4300 samples were

/zed to produce 11,000 measurements. This system proved to be effective in handling

d 200 samples per day in addition to the quality control standards. During the run quality control standards (QCS) were analyzed. The quality control effort represented

: 10% of the total analytical effort.

99
Quality Control Standards

For an analyst to have assurance that a measurement system is working properly when s aa using a manual measurement control system, he must analyze a bench standard having a "knot? Irene value and plot the result on a control chart. There is a possibility the analyst's resummon might be biased to fall within the control chart limits when the value is known. Therefotifasu for management to obtain unbiased estimates of the measurement errors, standards with va| sum unknown to the analyst must be measured in addition to the known bench standards. Actual nali measurement control is usually limited to administrative control and could be ignored by reae the analyst. In an automated system, only unknowns are required. The computer provides j immediate evaluation of the analyst's result. If the computer indicates the measurement Mont in control, the analyst may analyze process samples. If not, the computer locks out the lares method until corrective action is taken and an analyzed standard is in control. This syile 'O tem provides realtime measurement quality control. W

One of the objectives of the program is to determine each method's precision over ijiitha stated range. To do this, quality control standards are arranged in a sequence to cover i;e: the method's range. If st

The quality control chemist uses a computer program to rapidly log standards onto t'j disk and sequence them. This achieves the objective of covering the range of the method*~tis with an equal number of standards. Figure 2 illustrates how the computer prompts the chljcai ist for the information needed to generate an alternating scheme of standards in a sequel |s "(ji He has an opportuntiy to approve input before it is stored. By alternating sublevel samj"j)f Fel that are one or two standard deviations apart, the analyst never knows which subsample h- has at the level being tested.

Upon final approval, the computer prints out a table similar to the one in Table I

j the data are stored on the QC disk. This printout is used in the laboratory as a guide ;

T n ac labeling the ampuls containing the standard solutions. J ithod,

i id t( After labeling, the ampuls are placed in a sample storage area. One standard is at jm

lyzed each shift a method is used. All levels of the methodoS range are tested byfiequiiii, ( the analyst to take the lowest numbered standard from the available sequence and use a C; >feret stant aliquot size. It is analyzed according to the procedure in the method. Upon com- ji cu] pletion of the analysis, the analyst uses the computer to calculate his results. He instates

jfint

tlom

:i ani

all of the data requested by the computer..

Figure 3 provides a sample of the dialogue between the analyst and computer. The

analyst's input is underlined. He calls for the"DATA" program to calculate results of

measurement. In this illustration, on line 14, the computer prompted "QC OR SPL?" and 1 analyst typed, "SPL". The computer responded "NO SAMPLE ANALYSIS UNTIL A QC STANDARD I! RUN THIS SHIFT." The computer checks to see if an "in limits" standard has been analyze

within the past nine hours. If not, the computer locks out that method until a standar

analyzed that is in limits.

The calculated value is displayed for approval before being tested by the computer

If the result appears questionable, the analyst can reject the calculation and enter new

data. After approval, the result is compared with the reference value's acceptable ran' The computer responds whether the measurement is within limits or not. What the analys

does next depends upon the response. The three possible responses are shown in Figure s ^ A three standard' deviation limit has been established as the out-of-control limit. Res between the two to three standard deviation limit are flagged with an asterisk. If two consecutive standards are flagged, the method may be going out of control and correctiv

action is required. The analyst and his supervisor evaluate the situation and take the

necessary action. In cases where they are unable to determine a cause for out-of-contr

situation and cannot get the computer unlocked with the unknown standards, bench standa

of known concentration have been logged into the computer. Known values provide the

analyst with guidance as to the direction of his bias. Also, if there is a mismatch be

tween the reference values in the computer and the actual values, these standards can b used to unlock the computer if the results are within the control limits.

The <

Its pi

I

100
Surveillance

Surveillance activities are performed by all personnel. Upon evaluation of a stanss analysis, the analyst is given a message to indicate the state of control of the

rement system and appropriate action is taken. In addition to the analyst, the super

i monitors the quality control results. Standards Laboratory personnel daily monitor

feasurement activities of the laboratory. The computer is programmed to daily print

summary of all standards measurements made during the past 24 hours on a terminal,

quality control chemist exercises his surveillance responsibility by reviewing the

rements and initiating corrective action as needed.

-Control activities can be checked at any time by a "QC STATUS" program. Analyzed

'ards can be reviewed by method, acronym, standard number, or all methods. An example

fee "QC STATUS" using the acronym subroutine to check a method is shown in Table II. hemist specified status by the acronym 2U02 for a uranium f luorophotometric method.

2_ denotes a specific instrument. The computer listed the stored quality control stanthat have been analyzed on that instrument. The analysis, reference value, dif5ce between the two, standard deviation, result of the evaluation ("IN" or "OUT"), numf standard deviations, plus the analyst's initials, crew, and date are given in the

-s report.

This information was plotted on a control chart so measurement performance of the

lid can easily be viewed. Figure 5 illustrates a control chart depicting the results

-.e "QC STATUS" sheet plus subsequent data. The control chart reveals a problem on the

: >f February. One of the reagents was contaminated. After it was replaced, the method

back in control and there were no further problems.

Quality Control Standard Data Evaluation

1

E In addition to method performance surveillance, quality control data can be evaluated

;thod, instrument, analyst, or time period. The results of these evaluations can be

.ed to determine if there are any problems with the components of a measurement system.

h III shows the evaluation of one analyst's measurements for the f luorophotometric

od. when all the data for a method are analyzed, the analyses are grouped according

^ference value. The mean and standard deviation of the reported values at each subleval

Calculated. Equations are fitted to the data by regression. They are used to calculate

-nates of the standard deviation and provide bias corrections. These equations are in-

=d into a program which generates bias correction tables and estimates of the standard

itions of various levels of the method. These reports are used by laboratory superrs and Nuclear Materials Control personnel in evaluating measurement performance.

The equations are also included in the calculation programs for the appropriate analyL method. All measurements are bias corrected and reported with the standard deviation,

fe statistics provide reliability estimates for analytical measurements.

PROCEDURE FOR REPLICATE SAMPLES AND MEASUREMENTS OF NUCLEAR MATERIALS

; Replicate measurements are required for all measurement methods used to generate data

'nuclear materials control and accountability purposes. The replicate measurements are

1 to estimate the combined random error variance for sampling and analytical measures. Sample points within the plant that are covered in the replicate measurement prore include input, output, and inventory locations. There are two parts to this proceReplicate Sampling

Multiple samples are taken from accountability sample points. At least two samples

taken and measured using a density method to test for homogeneity. Analyzing duplicate

les provides assurance that the samples are representative of the nuclear materials in

vessel from which they were collected. Most of the sampling is performed in remote

ytical facilities equipped with master slave manipulators. The facility is divided

cells which are used for sampling and analytical measurement work. Needle block sam-

's are used to sample some 75 measurement points within the plant. Samples are analyzed

101
by a densimeter which has a remote head in one of the sample analytical cells. The instti ^ ment readout panel is located outside the cell. This equipment is currently being inter- j ^ faced with our computer system. After two samples have been analyzed, the results are tested to see if they fall within the predetermined limits set up for that point. If the do, the samples are accepted as homogeneous and further analyses are performed. If not,

additional samples are taken until they satisfy the requirements for replicate samples. » M

Replicate Measurements peed

. Hts o During a cold run using natural uranium, 20% of the samples were resubmitted for re-|tent

plicate measurements. They were randomly selected by the quality control chemist and sut mitted within 48 hours to different crews for analyses. These samples were distributed

over the measurement period. The samples submitted for replicate measurements were coded as quality control samples from the measurement point from which they were taken. They \> were analyzed for nuclear material content, density, and acid concentration. The replies]

result was compared to the original result and if the difference between the results ex—-P' ceeded the limits for the measurement point, a third bottle was analyzed. All data were , handled by our computer and stored for evaluation at a later date. Eventually, they werd

transferred to magnetic tape for long term storage. These data were used by Nuclear Ma-p

terials Control personnel to estimate the random error associated with the sampling and

analytical measuring activities of the measurement process. Figure 6 is an illustration

a final report generated by the computer for one of the replicate samples submitted as pi. quality control sample for a measurement point.

MASS AND BALANCE QUALITY CONTROL ACTIVITIES

The purpose of the mass and balance quality control segment of the program is to con with the requirements specified in 10CFR70.57 (b)(8)(i), (b)(12). Procedures are providi

for measuring and documenting mass measurement reliability. They provide data for randoi error estimates on weighing measurements. They are applicable to all mass standards and'j

balances used for the measurement of nuclear materials.

OUTSIDE MEASUREMENT EVALUATION PROGRAMS

The Safeguards Analytical Laboratory Evaluation (SALE) Program administered by the

United States Department of Energy's New Brunswick Laboratory is sponsored by the NRC an DOE. The goal of the program is to provide a means for laboratories to demonstrate thei

continued proficiency in safeguards measurements of nuclear material by periodic interlaboratory measurements comparisons. The Barnwell laboratory participates in the SALE p gram for the following reasons:

1. To demonstrate our analytical measurement capabilities.

2. To provide outside surveillance of our nuclear material measurement processes.

3. To provide an independent evaluation of our measurement activities.

4. To compare our measurement ability with those of other participants in the prog

ANALYTICAL STANDARDS AND REAGENTS

All standards and reagents used in the laboratory are prepared by the Standards Sec tion. This minimizes the error introduced by having a large number of people making the own reagents and standards for analytical chemistry methods. For nuclear materials measu

ments, standards are prepared having traceability to the National Bureau of Standards.

Most of these standards are packaged in ampuls and flame sealed as mentioned earlier. R gents and standards preparations are documented and dated. If they are suspected when a measurement system goes out of control, this provides a means of checking them.

SUMMARY

This program has been developed to include proven measurement control techniques,

innovations of modern technology, and requirements for good measurement reliability. Du

periods of high measurement activity, the quality control activities represent 10% of tr analytical effort. The computer system:

5,

102
Permits the use of a minimum mumber of standards,

Allows physical control of the measurement process by locking methods out,

Handles large quantities of data using modern record handling techniques*

Permits readily available surveillance of the measurement process,

Provides in-depth evaluation of the various components of the measurement process,

Bias corrects the measurements and calculates the appropriate standard deviation.

•ocedures for replicate samples and measurements provide random error estimates for

ts of the measurement systems. The other segments of the program assure us that our

jment activities are under control.

FIGURES AND TABLES

[. Computer printed table of quality control standard sequence.

[I. Computer printed program for obtaining the status of analyzed quality control

standards.

[II. Computer printed evaluation for an analyst's quality control standard

generated during the measurement period.

1. Schematic showing the components of the laboratory computer system.

2. Computer printout illustrating the login program for a sequence of quality control standards.

3. Computer printout illustrating the program for quality control standard data

entry by an analyst.

4. Quality control standard flow diagram for various responses by the computer.

5. Method U-F-l-B quality control chart illustration of plotted data from Table II and subsequent data. Analyst's initials and standard's identification number

are listed.

6. Illustration of computer printed final report of analytical measurements on a sample submitted as a quality control standard for replicate measurements.

103
Table I. Computer printed table of quality control standards sequence.

METHOD

QC STD # LEVEL SUB LEVEL REF VALUE SD

2U 501 1 1 1 .1 2U 502 1 2 1.2 .1 2U 503 2 1 13 1. 3 2U 504 2 2 11.7 1.3

2U 505 3 1 56 2.5

2U 506 3 2 58.5 2.5

2U 507 4 1 98 H U 2U 508 4 2 96 4.0

2U 561 1 2 1.1 .1 2U 562 2 2 11.7 1.3

2U 563 3 2 58.5 2.5

2U 564 4 2 96 4.0

DATA STORED

Table II. Computer printed program for obtaining the status of

analyzed quality control standards.

QC STATUS

? ACRONYM = 2U02

WJ (DENOTES INSTRUMENT)

2U02

STD # REF EXP BIAS SD

501 1.00 1.15 .15 .1 502 1.1 1.05 -.05 .1 503 13.0 15.8 2.8 1.3

504 12.7 15.2 2.5 1.3

505 56 62 6.0 2.5

506 58.5 66.5 8.0 2.5

90005 30.0 36.0 6.0 1.5

90006 30.0 29.0 1.0 1.5

IN/OUT DEVIATION C/A DATE

IN 1.5 E/JAL 27-JAN-78

IN -.5 A/CJA 31-JAN-78

IN 2.15 E/JAL 03-FEB- 78 IN 1.92 B/LAK 06-FEB- 78 IN 2.4 E/JAL 07-FEB- 78 OUT 3.2 E/JAL 07-FEB- 78 OUT 4.0 E/JAL 07-FEB- 78 IN -.33 E/JAL 07-FEB- 78

104
^able III, Computer printed evaluation for an analyst's quality control

standard s generated during the measurement period.

ACRONYM = 2U02

KNOWN REPORTED BIAS ANALYST # OF STD

2.8 2.98 -.18 B/TMS -.642

2.8 2.65 .15 B/TMS .535

.24 .24 0 B/TMS 0 .24 .18 .06 B/TMS 1.818

2.8 2.74 .06 C/TMS .214

16 14.64 1.36 B/TMS .85 .23 .29 -.06 B/TMS -1.818

15.5 14.52 .98 B/TMS .612

15.5 13.29 2.21 B/TMS 1.381

50 54.22 -4.22 B/TMS -=843

.23 .24 -.01 B/TMS -.303

.23 23.71 -23.48 B/TMS -711.515 5 SD (POINT REJECTED)

ONE POINT AT K = 16 R = 14.64

ONE POINT AT K = 50 R = 54.22

3 POINTS AT K = 2.8 MEAN =2.79 STD DEV = .170587

2 POINT AT K = .24 MEAN = .21 STD DEV = .424264E-1

2 POINTS AT K = .23 MEAN = .265 STD DEV = .353553E-1

2 POINTS AT K = 15.5 MEAN = 13.905 STD DEV = .869741

TOTAL NO. OF DATA POINTS = 11

TOTAL NO. OF LEVELS = 6

STANDARD DEVIATION EQUATION:

STD DEV = A*K~2+B*K+C WHERE A = .00024, B = .05061, C = .02697

BIAS CORRECTION EQUATION:

CORRECTION = A*K~2+B*K+C WHERE A = -.00181, B = .0732, C = -.0194

105
BALANCES

TITRATORS

DENSIMETERS

SPECTROPHOTOMETERS

GAS CHROMATOGRAPH

Figure 1. Schematic showing the components of the laboratory computer system.

106
QC LOG

ID? MM

03-FE1-78 11 : 02

? SEQUENCE

STORE DATA (Y OR N) ? Y

METHOD OF ACRONYM (M OR A)? M

METHOD = U-F-l-B

UNITS = PPM

NUMBER OF LEVELS = 4 NUMBER OF STDS AT EACH LEVEL =_2

HOW MANY CYCLES = 2 INITIAL SEQUENCE NUMBER = 501

LEVEL 1 Rl = 1J)

R2 = 1.1

SD = ^1

OK (Y OR N)? Y LEVEL 2 Rl = 13

R2 = 11.7

SD = 1^3

OK (Y OR N)? Y

igure 2. Computer printout illustrating the login program for a sequence

f quality control standards.

107
DATA

03-FEB-78, 10:45

CREW/ANALYST = E/JAL

METHOD = U-F-l-B

WHICH INSTRUMENT

ENTER 1 FOR JA, 2 FOR ORNL (HCLA) , 3 FOR ORNL (UF6) ? 2 CALIBRATE (Y OR N)? N LAST EQUATION COEFFICIENTS

ON 22-JUL-77 BY E/JMR

A = .644664

B = .859144

C = .179854E-3

CORR COEF = .999998

QC OR SPL? SPL

NO SPL ANALYSIS UNTIL A QC STANDARD IS RUN THIS SHIFT

QC OR SPL? QC

QC NUMBER? 2U00503

ENTER QC DATA: 03-FEB-78, 10:45

AQUEOUS SAMPLE (Y OR N)? Y

DILUTION (Y OR N)? N SPL ALIQUOT (ML)? _1 5_

MIBK VOL (ML)? 2 ORG ALIQ VOLUME (ML)? .1 INSTRUMENT READING? 421.5

CONC = 15.79 PPM

RESULTS OK (Y OR N)? Y *RESULTS BETWEEN 2 and 3 SD

Figure 3. Computer printout illustrating the program for quality control standard data

entry by analyst.

108
ANALYZE

QCS

st "*"

REPORT ANALYZE

*QCS SAMPLE

"OUT" or 2nd CONSECUTIVE

CORRECT PROBLEM

NOTIFY SUPERVISOR

ANALYZE

QCS

"OUT" "*"

ANALYZE

90000

^

SERIES QCS

NOTIFY

STD LAB

"IN" , RESULT IS WITHIN THE ±2 STANDARD DEVIATION RANGE.

"*", RESULTS ARE WITHIN THE ±2-3 STANDARD DEVIATION RANGE.

"OUT" , RESULTS EXCEED THE ±3 STANDARD DEVIATION LIMIT. THE METHOD IS STATISTICALLY OUT OF CONTROL AND CORRECTIVE ACTION IS REQUIRED.

Figure 4. Quality control standard flow diagram for various responses

by the computer.

109
BSAJAL

3 SD

6AJAL

2 SD

3AJAL 5AJAL

1AJAL 4AJAL 1AJAL

1 SD 4AJAL

8ACGA

2ACGA BSAJAL 2ACGA

-1 SD

-2 SD

-3 SD

7ALAK

CONTAMINATED

MIBK +

3ALAK

27- 31- 03- 06- 07- 10- 12-

JAN- JAN- FEB- FEB- FEB- FEB- FEB78 78 78 78 78 78 78

13- 14- 15-

FEB- FEB- FEB78 78 78

Figure 5. Method U-F-l-B quality control chart illustration of

plotted data from Table II and subsequent data. Analyst's initials

and standard's identification number are listed.

110
FINAL REPORT

SAMPLE 11157

Sample Point: N-34

Sample ID: QC 32

REQUESTED BY AS/JPC

LOG IN 15-SEP-77, 15:33

SAMPLING DATE

9-15-77

APPROVED BY JDS

LOG OUT

15-SEP-77, 23:10

SAMPLING TIME

15:33

COMMENT

U-VI REQUESTED AND RUN

SMALL AMOUNT OF SAMPLE DID NOT TITRATE ENOUGH

ANALYSIS RESULT UNITS

U 0.5 G U/L

H+ 3.623 N DN 1.1222 G/ML

AT 25 DEG C

SD(ISIGMA) METHOD

.03

.07

.0006

U-VI-l-A

ACID-VP-1-B

DN-DM-1-A

ANALYST DATE

C/LAK

C/LAK

C/LAK

15-SEP-77

15-SEP-77

TIME

18:10

20:06

15-SEP-77 20:20

Figure 6. Illustration of computer printed final report of analytical measurements on a sample submitted as a quality control standard for replicate measurements.

Ill
ACTINIDE DETERMINATION BY ROCKWELL HANFORD OPERATIONS' ANALYTICAL LABORATORIES

G. BURCH

Rockwell Hanford Operations

Richland, Washington

ABSTRACT

i

Lr

ijjoanic

A review of the procedures currently used by the Analytical Laboratories Department

of Rockwell Hanford Operations for the determination of actinide composition.

Methods used include extraction followed by alpha energy analysis, visible spectrophotometry, x-ray emission and x-ray diffraction techniques.

INTRODUCTION

Actinide analysis required of Rockwell Hanford Operations Analytical Laboratories Group

is done on several different matrices.

The first matrix has a high nitrate concentration; the actinides being present in the

hydrate metal nitrate form. The samples of this type are from the feed, waste and product

streams of the PUREX chemical separations plant. Samples vary in amount of nitrate present

(both as free acid and complexed with the metal ion), actinides present, and concentration

ranges. Types include 397-445 g/1 Th with 6 x 10" 5 - 2 x 10-2

g/1 233 U; <0.05 g/1 U with

3 x 10" 3 g/1 Np; 7 g/1 U with only trace amounts of other actinides; 238-506 g/1 U with

3 x 10" 5 - 8 x 10-3

g/1 Np; 10-26 g/1 U with 5 x 10" 2 g/1 Np; 2 x 10" 2 g/1 Pu and 2 x 10~ 2 g/1 Th; <.l g/1 U with 3 x 10-lt - 1 x 10" 3 g/1 Pu and others that have not been as well

characterized as yet. These samples are analyzed predominantly for Th, U, Np and some Pu. X-ray emission analysis is used to characterized solutions with U and Th concentrations 2 g/1 and greater. A visible spectrophotometry method is used for determining Th concentration in lower level solutions. An extraction/f luorimetric method is used to determine urar ium at concentrations lower than 2 g/1. Extraction and alpha counting are used in the ana- lysis of the isotopes 233 U and 237Np. Extraction and alpha energy analysis (AEA) methods a used to analyze for Pu content.

i Si pn The second matrix consists of wastes varying from a pH of 8 to solutions 8 M in hydroxide, - the matrix having been adjusted to maintain the integrity of the steel tanks that the solutions are being stored in until appropriate long-term waste storage methods are developed

The samples of this type were the effluents from the bismuth phosphate, REDOX, PUREX and

Plutonium reclamation chemical separation plants. These samples consist of four types.

1) Sludge - The insoluble metal hydroxides formed from the neutralization Kith NaOH. con- taining iron and aluminum as major metallic components with manganese, magnesium, chromium

actinides and other miscellaneous metals as minor constituents. When analysis of this samf

is required the sample is first centrifuged and the interstitial liquid removed. The solic are then washed with water several times to remove any water soluble components present.

The solids are fused with KOH or LiB02 and then dissolved with HC1 and water. 2) Tank Far

- The water soluble constituents of the waste tanks. These samples are combinations of the- constituents present in the following two sample types, found here in a more diluted form. : 3) Salt Cake - The salt left when tank farm material is put through an evaporation process

This consists mainly of the sodium salts of nitrate, carbonate, and aluminate with some sodium nitrite and other salts present in minor quantities. Actinides are present here as minor to trace components. When salt cake is analyzed it is first centrifuged to remove || interstitial liquid, the solids are then dissolved in water and heated. Any water insolubl^

solids that remain are fused with KOH, then dissolved using HC1 and water. 4) Interstitic

Liquid - The concentrated liquid left in the evaporation process. It contains dissolved

sodium nitrates, aluminates and other salts in a high caustic solution. Trace levels

IjMn

let MS

Iziii

'J 10 HI

! lea

itoniu

its,

112
tinides are found in these samples. These samples are analyzed for Pu and U content,

-raction/fl uorimetric method is used for determining U concentration. Either an ex- ion/alpha counting or an extraction/alpha energy analysis method (AEA) is used to mine Pu content.

vhird matrix consists of a wide variety of specimens from the plutonium processing plant.

; samples include plutonium nitrate solutions, oxide and metal. Scrap from engineering

es, samples from hood and process areas, process stream samples, and solid waste samples

ill a part of this sample type. These samples are analyzed for plutonium content both

;afeguards and accountability. Physical plant items, such as gloveboxes, hoods, ducts,

n'pes are also in this third sample type and are analzyed in situ and may be dismantled

)lutonium recovery. Amperometric and visible spectrophotometry methods are used for

mining Pu content in the plutonium nitrate, oxide and metal samples. Alpha counting

;ed to determine Pu levels in process streams and waste samples. Non-destructive

n'ques are used in the Pu determination of the rest of the items. It should be noted

when the non-destructive techniques are used it is the item itself that is analyzed

Dst cases and not just a sample of that item.

g x-ray diffraction techniques the fourth matrix is analyzed for actinide composition,

ganic powders and masses are analyzed to determine the presence of actinide compounds

to identify the chemical forms of the compounds present.

in Solutions of High Thorium Concentration

[matrix for this sample type is that of high nitrate as described previously. The first

I of this procedure therefore is a calcination process using HC1 to destroy the nitrate.

uranium and thorium form nitrate complexes which extract into the organic phase. The

rds left from the calcination are then dissolved with 9 M HC1 and the uranium chloro

!«lex is extracted into 5% TIOA (tri-iso-octylamine) in xylene. This accomplishes the

.. extraction of uranium from thorium since thorium does not form anionic complexes with

chloride ion. Refined separation is then further accomplished by stripping the uranium

» 0.1 M HN0 3

. Although separation is complete at this point this solution corrodes the

ia discs used for counting the samples and gives erroneous results. To separate out

mounting interferences the pH of the samples is adjusted to 3 and the uranium extracted

: d 0.5 M_TTA (thenoyl -tri-fl uoroacetone) in xylene.

'aliquot of this extraction is then mounted on a one inch stainless steel alpha disc and

Dorated to dryness on a "cold spot" heating stove or a wire heater. The disc is then

Tied to reduce self absoption and fix it to the disc. The sample is then counted on an

; na proportional counter (APC). Recovery on eight standards (5.046 x 10 7 dpm/1) run over

eriod of two months showed a mean recovery of 93.5% with a standard deviation of 3.1% at

95% confidence level.

Np in High Level Beta-Gamma Samples

s method is seldom used in our laboratory at the present time, but is included for comteness. The sample is adjusted so that the acid content is 4 M. Ferrous sulfamate and

razine are used to reduce neptunium to the (IV) state. The actinides are extracted

ng 10% (by volume) TIOA (tri-iso-octylamine) in xylene. The actinides are then stripped

0 1 M HC1 and ferrous sulfamate and hydroxylamine-hydrochloride added; the solution is m heated. At the end of this process neptunium is in the extractable (IV) state while

itonium is in the inextractabl e (III) state; uranium is present in the inextractable (VI)

ite. The Np is then extracted into 0.5 M TTA (thenoyl -tri-fl uoroacetone) in xylene,

aliquot of this is mounted on an alpha disc, evaporated to dryness, fired and counted

ng an alpha proportional counter (APC). No accuracy and precision data was available.

'Np in Solutions of High Uranium Concentration

using this procedure the plutonium to neptunium ratio of the sample is in the range from

to 10. The hydrogen ion concentration of the sample is first determined then hydroxylEXTRACTI0N AND COUNTING METHODS

113
amine hydrochloride, ferrous sulfamate and deionized water are added. The hydrogen ion || centration is then adjusted to 0.9 H using HN0 3 or NaOH. By these steps the neptunium if! reduced from the (V) and (VI) states, in which it is most prevalent in aqueous solutions,;

the (IV) state which is extractable. The plutonium is reduced to the (III) state which i; inextractable; the uranium is left in its inextractable (VI) state. The neptunium is ttf) extracted into 0.5 M TTA (thenoyl-tri-fluoroacetone) leaving Pu (III), U (VI), Fe (III)j

sodium, aluminum and potassium ions which interfere with counting due to alpha absorption;

The neptunium - TTA is then washed with .8 M HN0 3

, mounted on an alpha disc, evaporated!;

dryness, fired and counted by an APC.

In 1972 using seven pieces of data an average recovery of 88.3% was obtained with the pn|

cision of a single observation 29.8% at the 99% confidence level. Standard recoveries r| lately have yielded only about 70% recovery.

Pu From Other Alpha Emitters and Fission Products

This procedure is the most commonly used of the two available for Pu analysis in our latij

oratory. The hydrogen ion concentration of the sample is first adjusted to 2-3 M, then j ferric nitrate, hydroxylamine hydrochloride and aluminum nitrate are added. The solution

is then heated. The heat and acid are used to destroy any plutonium polymer present wlii- the aluminum nitrate ties up fluoride ion and the ferric ion and hydroxylamine reduce thjl Pu to the (III) state by the following mechanism:

NHoOH; + Fe+3 < Fe+2+ H,0 + 2H

+

+ 1 No \ 2

Pu(III, IV, VI) + Fe+2* Fe+3+ Pu (III)

! '.0 CO After cooling, the acidity of the sample is adjusted to 0.8 - 1.0 M. Sodium nitrite is tj

fj xe used to form nitrous acid which oxidizes the Pu to the extractable (IV) state. The plutonium is then extracted using 0.5 fl TTA in xylene and an aliquot mounted on a one inch

stainless steel alpha disc, dried, fired and counted on an APC.

s dat

k Iiral

Itri

If samples containing high 95 Zr- 95 Nb are analyzed the mounts will have very high beta-ga

activity which will lower the beta threshold of the APC down onto the alpha plateau resu

in an alpha breadkown. If 95 Zr- 95Nb is present the Pu (III) is stripped into 8 M HN0 3 This strip will recover >99% of the Pu and -2% of the 95 Zr- 95 Nb. An aliquot of this phaljjy is then mounted , dried, fired and counted as described above. Iron (III) will also ex tract into the 0.5 M_ TTA in xylene and relatively large quantities (1 mg) can cause low

results because of alpha absorption. Am (III) and Cm (III) do not extract, however Np,

(

will extract and its contribution to the alpha count is corrected through alpha energy

analysis (AEA). An AEA is also run whenever the isotopic content is questionable since

small amounts of relatively short lived isotopes, particularly 238 Pu, can make a fairly

large difference in the sample alpha activity. Many of our samples are routinely report

as activity (or count rate) per unit liquid volume, however, these results can be convert

to grams using the appropriate isotopic correction factor. In most cases the concentrat

of Pu in these fission product waste samples is relatively low and may be reported as 1C 239 Pu. Because of the small amount of shorter lived 2I+0 Pu present in all Hanford waste

results using this conversion method will always be biased high since average AEA detect

have a resolution of 25 keV @ FWHM and thus cannot distinguish 240 Pu from 239 Pu. This fc is likened to a "worst case" situation, and usually is not too big a problem. Small arnc of 238 Pu can increase the alpha activity by orders of magnitude and result in a very hig]

bias (e.g 20 wt.% 2lf0 Pu will increase the uncorrected answer about 1 h times, while an e alent amount of 238 Pu will increase it 56 times ). If a result based on 100% 239 Pu is 0. or greater then an AEA is run and the results corrected for any 238 Pu present.

This method has been found to give, under controlled conditions, an accuracy of 95.9% +

at the 95% confidence level. As a check on the method a standard plutonium solution is | routinely analyzed containing Pu in a matrix similar to the samples being run. For the

month of Janurav 1978 five standards were run giving an average recovery of 96.6% with e standard deviation of 8.2%. At the 95% confidence level the precision of a single obser

vation is 22.6% with the precision of the average at 10.1%.

114

flu j

Hi
um in Caustic Waste Sample

"•ocedure is a new one and still is undergoing refinement. The procedure now gives

•out 65% recovery for Pu, however since a tracer is used this helps reduce the problems

-/ recovery causes. In order to do this analysis the concentration levels of the folare needed in order to adjust the acid concentration: hydroxide, aluminate, carbonate

)sphate. Carbonate and phosphate are neglected if they are known to be present in

Lrations considerably less than the OH" and A1(0H)^. A

236 Pu tracer is added to the

I and allowed to equilibrate. The sample is then acidified with concentrated HN0 3 to

iity of 4-5 M. Urea is added and the sample is heated to destroy any nitrous acid

| (nitrous acid will consume reducing agents used in later steps.) (An excess of urea

s avoided as it will consume nitrite that is added later to oxidize Pu (III) to Pu Ferric nitrate and hydroxylamine hydrochloride and heat are then used to reduce

) and (VI) to Pu (III); the heat will also destroy anv polymer that may be present,

ium (V) will be reduced to Np(IV) and U and Am will be unaffected. After the solution

led sodium nitrite is added to oxidize the Pu (III) to Pu (IV). The Np (IV) is not %d since the acid concentration is > 4 H; Am remains in the (III) state. Pu (IV),

with any UOt4and Np (IV) is then extracted by adding an aliquot of 30% (by volume)

't-336 in xylene. Two batch contacts are made and then a scrub with 8 M nitric is done.

jeous phase can then be used in an americium determination.

n and some salts are then stripped from the organic using 10 M HC1 . The organic then

dium carbonate and xylene added to it so that the Aliquot is diluted to 10% to assure

itative stripping of the actinides into the aqueous phase. Following a wash, conceni HC1 is added to destroy the carbonate. The solution is then evaporated to dryness

i some sodium bisulfite has been added to prevent baking of the actinides on the vial

drying. After cooling, the electrolyte, sodium sulfate, is added and the pH adjusted

'amonium hydroxide until a pH of 2-2.3 is reached. A platinum electrode is then used

ctroplate the actinide onto an electropol ished disc. Ammonium hydroxide is added just

to completion of the electroplating; this causes the hydrous oxide of the actinide to fixed to the disc. The disc is rinsed with ethanol , dried and then counted on an APC

1 AEA methods. The 236 Pu spike recovery is then used to correct for any procedural

1 that have occured. Certain anions, such as sulfate, oxalate, phosphate and organic

inc agents nay inhibit the extraction by complex formation, resulting in lowering of

ural recovery. However, since a 236 Pu spike is used and presumed to be in equilibrium

he sample these interferences can be corrected for.

from the aqueous

in xylene. As

is used to check

s being continued on finding an adequate method for Am determination

obtained in this procedure after the extraction with 30% Aliquot-336

! s date the methods tested give only a 10% recovery when a

21+3 Am pike

'.ural recovery.

NON-DESTRUCTIVE ASSAY

1 metric

'lantity of plutonium in a package is determined by measuring the heat output of the

; package and quantifying this amount using a standard. The isotopic distribution is

lined via a gamma scan since the heat output of the sample is dependent on the relative

! :s of isotopes present. The specific heat/isotope is then used with the isotopic dis-

:ion and the heat output to determine grams of Pu. Current measurement error is estito be +.5 to 4.5% relative standard deviation depending on isotopic composition.

hunting l

anuria detectors and associated electronics which record gamma radiation in the 375-450

egion are used to determine plutonium quantities. These systems are used in fixed

ions to determine Pu amounts in items such as half liter polyjars containing wet sludge,

systems are also used in a portable mode to determine quantities of Pu in such things

ilding pipes, ducts, hoods an'-' gloveboxes. The plutonium content is determined by renent of the 239 Pu gamma ray complex emitted at 414 KeV, correcting for compton

ering and background, and comparing to calibration standards. In situ measurements of

nium in several gloveboxes were found to give agreement within 16% with actual amounts
Plutonium Isotopic Analysis

Items analyzed on this system must meet the requirement of homogeneity throughout, sinct density of the item causes the low energy radiation to be absorbed so that the detector

counts only the outer layer of sample facing it. The system uses a high resolution ill

sic germanium detector to collect a gamma spectrum covering the energy region from 20 k 420 keV. The spectrum is analyzed by computer. The computer analysis involves the cori struction of a relative counting efficiency curve for each item by comparing known bran ratios of various 239 Pu and 2kl Pu peaks with their observed intensities and then normal

calculated analytical peak areas. Actual isotopic ratios are calculated using closely eCor spaced peak pairs with the subject isotope activity being compared to the activity of %f$o or 241 Pu in a nearby peak.

Segemented Gamma Scan Assay

A Ge(Li) detector is used to measure the 414 keV gamma complex used to determine 239 Pu content in packages. This system is not used on items containing <_ 1 gram Pu. The san is rotated at a constant speed during counting to reduce the effects of radial inhomogc

within the item. Correction for variations in density are made by simultaneous measun [} of a 75 Se (400.7 keV) gamma transmission source mounted directly opposite the detector,.]^, A

133 Ba (356 keV) gamma source is mounted on the detector housing and is used for dead

and pulse pile-up corrections. The item is scanned in individual vertical segments an(

the 239 Pu values are computed on a segment-by-segment basis using the appropriate calit

tion factors and correction factors for density, dead time, and pulse pile-up. The sun fey the individual segment results gives the 239 Pu value for the package. This value is t\ used with the 2

'+0 Pu isotopic content estimate to calculate the total plutonium content,

measurements on a 40 g Pu standard gave an average recovery of 100.15 + 4.5%. For six oj measurements on a 160 g Pu standard an average recovery of 98.5 + 1.45% was obtained, ijite

ii

sta

AMPER0METRIC METHOD met

i w! t The amperometric titration method is used to determine the plutonium content in pluton"

nitrate solutions, plutonium oxide and plutonium metal. The first step in this method

a fusion with sodium bisulfate for the oxide and metal samples. The samples are then < solved in "Q" water to give a plutonium concentration of 5-12 mg/rnl . The nitrate solu'

is diluted to give a similar amount. To the sample H 2 S0 1+ is added and the sample stin

While the sample is stirring an excess of argenic oxide (AgO) is added to convert all |K

(IV) to the Pu (VI) state. The excess argenic oxide is then removed by heating. A w|||( excess of ferrous ammonium sulfate is then added to convert the Pu (VI) to Pu (IV). T^ excess ferrous ions are then back-titrated with potassium dichromate, using an amperomi

endpoint detection method. The plutonium content is then calculated from the amount o ferrous ions needed for complete oxidation of the plutonium. The estimated relative s\

dard deviation for the analysis of plutonium oxide is 0.1% with an average accuracy fob measurement of 100.03%

FLU0RIMETRIC METHOD

This method is used for uranium analysis in concentrations less than 2.0 g/1 . and has

us^d periodically for many years. When free from quenching agents such as Cr 3

, Cr+6

, Fe 3

, Al +3

, Na+1

, Mn +?

, and HN0 3 in process solutions, an aliquot of the sample is mou directly on a platinum dish, dried, fused with NaF - Li F (98%-2%) and read on a calibr.

fluorimeter. If quenching agents are present the U (VI) is extracted into methyl isob

ketone (hexone) from an aqueous solution heavily salted with ammonium nitrate or alumi

nitrate. No accuracy and precision data was available.

iitc

VISIBLE SPECTR0PH0T0METRIC METHOD

Thorium

This method is used for determining Th concentration in samples with concentrations le than 2 g/1 Th. The sample is first treated with HC1 and sodium nitrite to convert all to the (IV) valence state. 10% TI0A (tri-iso-octyl amine) in xylene is then used to exj

Zr, Hf, U (IV), Pu (IV), and Np (IV) from the Th. These metals all form interfering e

116
ins with the complexing agent. The extraction is done a minimum of two times and

a third time if the Pu/Th ratio is greater than 500. An aliquot of the aqueous

ning the Th) is then adjusted to an acidity of 5-9 M with con. HC1 , and sulfamic

- ded to destroy any excess nitrite present. Arsenazo III, which is used for the

h'ng agent, is sensitive to both oxidizing and reducing agents so it is necessary to that neither are present for the complexing process. An aliquot of 0.05% Arsenazo

: 8-dihydroxynaphthalene-3,6-disulphonic acid-2,7-bis-[ (azo-2) phenylarsonic acid])

I mixed with the sample. The absorbance of the sample is then read against a reagent

|t 660-665 nm, and the sample absorbance used to calculate the Th concentration. A •ecording double beam spectrophotometer is used for this determination. About 1 yg/

»f solution may be detected by this method. The best working range is 2-20 yg/10 ml.

lard is run with each batch of samples run through the method to check for recovery,

mdards have been giving a 99+7% recovery.

f)rk has been done on a spectrophotometry method for the determination of uranium when

: in concentrations less than 2 g/1 using Br-PADAP [2-(5-bromo-2-pyridalazo)-5-diethDphenol]. However, demand for this analysis is small so the final work has not been

fed.

ium

f-2thod measures the concentration in g/1 of Pu in aqueous samples or solutions of dis-

' metal. Hydroxyl amine hydrochloride and heat are used to reduce the Pu to the (III)

'e state. The blue solution is cooled to room temperature and then diluted to volume,

'sorbance of the blue sample solution is measured at 565 nm relative to that of a blue

filter on a double-beam spectrophotometer. The most accurate measurements ere

ed when the solution has a concentration between 11.0 and 11.5 g/1 Pu. The precision

s method has been reported as +0.6%, no accuracy data was available. This method is

red to alpha counting techniques for plutonium process solution samples >_1 00 g/1 Pu.

X-RAY EMISSION ANALYSIS

L

m concentrations are determined annually on solutions in storage tanks as part of the

|SS Nuclear Material Inventory. Solutiors with U concentrations 2 g/1 or greater are

ed using this method.

I|m concentration is determined in aqueous solutions using yttriun as an internal conftandard. The internal control -standard is used to compensate for variations in vol-

: emperature, density and bubble formation in the liquid mounts. A constant amount is to all samples and calibration curve mounts to provide this correction. Calibration

e using the linear relationship that exists between uranium concentration and the

of uranium counts to yttrium counts. Two calibration curves are prepared covering two

•ent concentration ranges as Table I shows.

TABLE I

Curve

mg U Mounted

mg Y Mounted

U Concentration

Range Covered

A 2 to 12 2 2 g/1 to 60 g/1

B 20 to 77 20 25 g/1 to solubility

1 imit

117
Sample aliquots and yttrium aliquots are then chosen according to the concentration ra of the sample. The aliquots are pipetted into 5 ml volumetrics and then diluted to vo with 1 N_ HN0 3

. Two aliquots of this mixture are then mounted in x-ray sample cups tha

covered with mylar. The samples are then counted on an energy dispersive x-ray emissic

spectrometer using a molybdenum tube with molybdenum filter. The FWHM (full width hair

imum count) integration area is then used to determine the number of counts under eachj

On our system we commonly use the areas 13.54 to 13.78 keV for U, and 14.38 to 15.12 kf 1 '^ Y. A linear regression analysis is then used with mg U as the abscissa and U/Y ratio | ordinate. The concentration ranges of the two curves shown in Table I appear to be lir over their ranges, however extrapolations of either curve outside it's stated range ha\

been found to give incorrect results. Results obtained on a standard using the low cut have given an average recovery of 96.8% + 2.0%, while results obtained using the high

have given an average recovery of 97.7% + 1.4%.

H tt

Thorium

nun

3 r ?

Because of fairly stringent accountability requirements for this group of samples th i s P

|[

ique has been investigated quite extensively. When this x-ray method for determinatioi

Th concentration levels was first developed a dry mount was used, however a dry mount v found to be more susceptable to error, had a higher detection limit and required a lon( ;:fll

5 sample preparation time. Like the uranium method just described this one also utilizer

yttrium as an internal control -standard , and the sample mounts are prepared in the sam< We use calibrated pipets and volumetrics in working with these samples although using 1' calibrated volumetrics will show little effect on the standard recovery in 98 out of H

samples. Whenever the same pipet can be used between samples, standards and calibrate

curve mounts we do this as it simplifies the calculations. One pipet can be used to a< yttrium to all of the vials, similarly one pipet can be used in mounting the samples, j though here again small volume differences will have a negligable effect on recovery,

our initial sample characterization we ran a full calibration curve covering a range 0 to 238 mg Th/5 ml with each set of sample mounts. However, since that time, we have f

that it is only necessary to run a two point calibration curve because of the linear c tion range and the close concentration levels found in our samples. Samples are count!!*

the energy dispersive x-ray emission spectrometer using the molybdenum tube and filter

the following peak integration areas; 12.88-13.12 keV for Th and 14.84-15.08 keV for Y have found it unnecessary to do a background subtraction; neglecting the background ha! than a half a percent effect on the recovery obtained. When the calibration curve is an average recovery of 99.2% +_ 1.1% at the 95% confidence level is obtained. With reel data obtained using the two point calibration curve we got an average recovery of 100. + 0.54%, however this data was obtained on a much smaller number of data points than t for the calibration curve and so gives a misleading indication. We feel it does show, ever, that with our sample type we have obtained essentially as good of results with t point curve as with the full calibration curve, which reduces the amount of work neede

cal ibration.

liif

ay po

ui(iu

lis

X-RAY DIFFRACTION ANALYSIS

As mentioned previously x-ray diffraction analysis is used in our laboratory as more 0 qualitative method than quantitative, however I mention it here because some of the te iques we use may be of use to others. Specifically I would like to mention some of th

mounting techniques that have been developed that have greatly improved our results.

Oxide compounds of the actinides have a well-known tendency to creep and become easily,

borne. Because of these characteristics and our need to keep our x-ray instruments fr

radioactive contamination we have developed a mount for use on a diffractometer that i minimum expousure to personnel preparing them, complete sample containment and essenti

no background effects on the diffraction scan. When we receive our samples they have

ready been ground to a fine powder state; using a transfer pipet, resin is added to th

powder in the flask. This involves little personnel exposure since the flask can be s a block of shielding material and no close physical contact is necessary in using the

A few drops of fixer are then added to the flask and the contents stirred using the tr pipet. Sample can be drawn up into the pipet as an aid to mixing and a check of it's

pleteness. A sample mount is prepared by placing some mylar on a glass slide and coat

it with releasing agent; a glass ring is then placed on the mylar. Previous mounts usl

118
3 elite rings for the mount but we have found that glass has a minimal background affect

; also allows you to see through the side of the mount to check and see that the sample

=:ure is without air bubbles and other complications. The sample/resin mixture is then

nn up into the pi pet and pipetted into the glass ring. The sample is encased by the

[in in this process so there is no danger of it's becoming airborne. The pipet and sample

rl are then disposed of as radioactive waste. The sample is allowed to dry, which takes

jt 24 hours. When the sample is dry it is lifted off the mylar which is disposed of (we

2 found this piece to have quite a bit of contamination on it.) We then wrap the mount

ig two layers of Kapton X-ray Film (Spex Industries) which we have found to be the best

fjn for x-ray diffraction use. The film is sealed at the back with tape. Depending on radioactive level of the sample we then use either our hands or long tweezers to place

mount in the diffractometer . We have deliberately tried to break this mount by dropping

and bumping into objects with it; as of yet we have been unable to do so, hence we feel

mount is safe even if fairly roughly handled. If the wrap were removed from the sample

2 smearable contamination could occur, however, this amount would be small and confined

a small area; it is highly doubtful that any of the actinide could become air-borne. An- advantage of this mount is that the sample surface is flush with the front of the

: ;le mount so there is no need to worry about shift in d-spacings obtained. Using a nt prepared in this manner of a-Si0 2 no background effects were observable other than a aring of count rate as compared to a mount with a similar amount of the powder. We cur- tly use the glass ring/Kapton X-ray Film combination on all of the diffraction mounts

•lyzed in our laboratory. We use two glass rings cut so that one (or two if a radioactive

ole is to be analyzed) layer of Spex wrap will fit between the rings giving a flush

•face. We have been unable to observe any background effects caused by this mounting

hnique on any of the sample types that we use in our laboratory.

i diffraction scans are needed on very small quantities of radioactive sample we use an ay powder diffraction camera mount. This mounting technique was developed by combining

iiniques shared with us by Mr. P. L. Wallace of Lawrence Livermore Laboratory and Mr. M. C. ihols of Sandia Laboratories. A very small amount of powder is placed on a glass slide

: a drop of rubber cement/benzene mixture is added. A second glass slide is used with a ^cular sliding motion to form the powder/glue mixture into a small ball. Because such a 01 sample amount is used the radioactivity level is low enough that no remote manipulation

necessary. A gel strip or rod moistened at one end is then used to pick up the ball and

: two are dropped into a glass capillary. The safest and easiest method to use in handling

glass capillary is to insert it through a cork stopper into a test tube. The capillary

l;ide the test tube is easier to hold, will not become contaminated and can still be seen.

!, glass capillary is then sealed and used in the powder diffraction camera. We are instigating the possibility of using the rod/sample ball directly in a Gandolfi camera (here

will use a glass rod with glue on the end to hold the ball.) Using this technique the

i;m mounting, as well as the other operations will have to be performed in a hood.

samples are analyzed on the diffractometer and camera using standard techniques. We use

opper tube on our diffraction instruments. We have two horizontal diffractometers , one

which is set up on the medium focus so that a camera can be used simultaneously with it.

other diffractometer we have set up with a Si (Li) detector and a single channel analyzer

rthe fine beam focus. This instrument gives particularly good resolution for powder

; fraction studies, with it we can resolve the a-Si0 2 quintuplet found at 1.37-1.38 d-spacI

SUMMARY

(overview is given of the methods used for actinide characterization by the Analytical

'oratories Department of Rockwell Hanford Operations. Brief explanations are given of the

• enteen methods used in the analysis of widely different sample types containing the actdes in concentration levels ranging from just above background to the solubility limits.

J

119
ACKNOWLEDGMENTS

I am grateful to Chuck Clark and the chemists of the AL Department for their help in data

collection and review. Gratitude must also be given to Wayne H. Sant who encouraged me to start and complete this project. Secretarial and editorial assistance was efficiently

carried out by Linda A. Leemhuis.

REFERENCES

1. C. H. Kindle, In Situ Measurement of Residual Plutonium , ARH-SA-248, Atlantic Richfielc

Company, Richland, Washington, June 30, 1976.

2. S. J. Johnson and W. I. Winters, Actinide Analytical Program for Characterization of

Hanford Waste , ARH-SA-296, Rockwell Hanford Operations, Richland, Washington,

October, 1977.

For more detail on a particular procedure, please contact the author.

120
X-ray Diffraction Sample Holder for Radioactive

Samples or Samples that React with Air or Light

by

H. W. Dunn

Oak Ridge National Laboratory, Oak Ridge, Tennessee

ABSTRACT

A very versatile sample holder for an x-ray diffractometer will be described. It provides double containment

while in operation on the diff Tactometer and triple containment while being transported from a hot cell or dry box to the diffTactometer and back to a disposal area. This cell

is suitable for fairly high alpha activity or mild betagamma activity. It is also good for materials that react

with air or light. Either solid or liquid samples can be

used.

KEYWORDS: X-ray diffraction, analysis of radioactive

materials, transporting radioactive samples,

handling radioactive materials

INTRODUCTION

We occasionally get requests for x-ray diffraction analysis of radioactive materials,

t our equipment is located in a building that has no facilities for handling radioactive

terials . Therefore, any such materials that we handle must be at least double contained

FACILITIES FOR HANDLING SPECIAL SAMPLES

We have designed a sample cell that allows us to handle special samples where activity

involved. An exploded view of the device is shown in Figure 1. It is made of stainless

eel with thin plastic windows cemented in place.

Figure 1, Part 1, shows the cell in which the sample is placed. It is introduced

rough the slot "B". The thin plastic windows, "A", allow the x-rays to be transmitted

tth very little loss. After the sample is introduced, usually in a hot cell or dry box,

te cell is sealed with plastic tape. The sample cell can then be moved to another area

;d decontaminated. The sample cell is then inserted through slot "C" and seated in oove "D" in the base unit, Part 2. The slot and groove are machined for a slip fit

as to hold the cell in proper alignment when it is placed on the dif f ractometer

.

The containment unit, Part 3, is then inserted over the base unit and aligned by the

ot "I" in the containment unit and pin "F" in the base unit so that the primary and

ffracted x-ray beams pass through the plastic windows on the cell and containment unit.

|e assembly is locked together by means of the knurled screw "K". A thin plastic window

[", "0" ring "E" in the base unit, and "0" ring "L" on the underside of the knurl screw

" completely seal the assembly giving double containment to the sample. Since the

ndow on the containment unit is thin, it could be punctured while in transit from the

t cell or dry box to the dif fractometer . To prevent this, a carrying unit, Part 4,

ips over the containment unit and is sealed by the "0" ring ,: J". These two units are

eked together by the knurled screw "P" that contains an "0" ring "D" on the underside

'perated for the U.S. Department of Energy by Union Carbide Corporation under contract

-7405-eng-26 . Research supported by the Office of Basic Energy Sciences.

121
to complete the seal. The threaded hole "N" does not go completely through the metal.

This gives triple containment while the assembly is being transported from the hot cell

or dry box to the dif f Tactometer . A track "R" is machined into the shaft head of the

dif fractometer , Figure 2. A tight slip fit with the matching track "G" on the base

unit and the stop pin "S" accurately position the sample on the dif fractometer . After

the assembly is placed on the dif fractometer , the carrying unit is removed and the

sample is examined to be sure no spillage has occurred. If not, the sample is then

ready for analysis on our Philips x-ray diffraction equipment.

The signal to background ratio is improved considerably if the diffracted beam

monochrometer is installed on the dif fractometer . If much beta-gamma radiation is

present, it would also help to install some additional shielding.

Since the sanples can be loaded and sealed in a dry box containing an inert

atmosphere, this assembly can be used for samples that react with oxygen or moisture

in the air. If an opaque material is used for the cell windows, this same system

can be used for samples that are affected by light. Slurried samples or liqud

samples can also be used.

After the analysis is complete, the carrying unit is replaced and the complete

assembly is removed from the dif fractometer . It is then carried to a disposal area where the cell is removed, cleaned, and made ready for the next sample.

By acceptance of this article, the publisher

or recipient acknowledges the U.S.

Government's right to remain a non-exclusive,

royalty-free license in and to any copyright

covering the article.

122
PART ••

PART 3

PART 1

PART 2

123

Feasibility Study of a High-Precision XRF System

for Analysis of Solutions and Slurries

C. R. HUDGENS and B. D. CRAFT

Mound Facility*, Miamisburg, Ohio

ABSTRACT

Wavelength dispersive X-ray fluorescence (XRF) is uniquely applicable to the

analysis of dissolver solutions because of its high immunity to the intense gamma

emissions of the solutions, its inherently high signal-to-noise (S/N) ratios, and

its adaptability as an on-line, tamper-resistant, accurate analysis of both dilute

and concentrated plutonium in dissolver solutions.

The taking of aliquots and transfer of samples would be eliminated by placing

the system on-line. Sampling thereby would become total: during the progress of

of the analysis the entire contents of the dissolver tank—or its logical equivalent, a large, thoroughly mixed fraction of its contents —would be circulated

through the sample cell. Internal standards would be used for realization of the

highest accuracy of analysis. The addition of standards would impose no complications to subsequent chemical processing.

The use of monochromatic exciting radiation would further improve the S/N

ratio, and, by eliminating useless radiation, would reduce photochemically formed

gas bubbles which perturb sample geometry. For rapid analysis a rotating target,

high-power X-ray generator would be necessary.

Control and monitoring of the equipment, the analytical process, and data processing would be done by a minicomputer for error-free operation, and for prompt

detection of all but the most sophisticated attempts at tampering with the analytical process.

The specifications of the system components have been conservatively estimated so that the r.m.s. variation of the entire system during a single determination would not exceed 0.20%. Counting statistics is considered separately

because it is a function of fluorescent intensity, and therefore may be limited

by available time for dilute solutions. For concentrated solutions with strong

fluorescent emissions the total variation of a single determination could approach

the 0.20% level. By using the usual strategy of repeated determinations, any

arbitrary precision could be reached.

The performance specifications have been written with awareness of the

capabilities and limits of modern technology. The capabilities are exploited

and the limits are not strained. The conservative designs thus permitted will

allow the construction of an XRF system of high reliability and advanced capability .

KEYWORDS: Nuclear safeguards; X-ray fluorescence analysis; rotating target

, X-ray tube; X-ray monochromator ; X-ray liquid and slurry sample cell; highprecision, wavelength dispersive X-ray spectrometer; analysis of solutions

and slurries

id Facility is operated by Monsanto Research Corporation for the U.S. Department of

;y under Contract No. EY-76-C-04-0053.

125
INTRODUCTION

Safeguards needs for on-line, tamper-resistant, accurate analysis of both dilute and

concentrated uranium and plutonium in dissolver solutions 1 have been documented. Wavelen,

dispersive XRF is uniquely applicable to the analysis of dissolver solutions because of i

high immunity to the intense gamma emissions of the solutions, and because of its inheren

high signal-to-noise (S/N) ratios. The use of monochromatic exciting radiation would fur;

improve the S/N ratio, and would largely eliminate photochemical gas formation which is a, analytically detrimental effect of the usual white exciting radiation. Monochromatizatio

does incur power losses, however, and for rapid, precise analysis, a very high power X-ra

generator of the rotating variety would be necessary.

The taking of aliquots and transfer of samples for XRF analysis would be eliminated

placing the system on-line, effected by installing the XRF sample cell in a recirculating

loop with the dissolver tank. In this mode, sampling errors would be eliminated, because

during the progress of the analysis the entire contents of the dissolver tank—or its log

equivalent, a large, thoroughly mixed fraction of its contents —would be circulated throu

the sample cell. In addition, undissolved material in the form of a slurry would be assa along with the dissolved material. Internal standards would be used for realization of t highest accuracy of analysis. The addition of suitably chosen standards would impose no complications to subsequent chemical processing.

Control and monitoring of the equipment and of the analytical process, and data proc

sing, would be done by a minicomputer for error-free operation, and for prompt detection

all but the most sophisticated attempts at tampering with the analytical process.

EQUIPMENT

is

: i

ilf

Sift

wii

(it:

The specifications of the individual components described in the following pages ha\

been conservatively estimated so that the r.m.s. variation of the entire system during a single determination would not exceed 0.20%, exclusive of counting statistics. The latt

is considered separately because it is a function of fluorescent intensity, and therefore

be limited by available time for dilute solutions. For concentrated solutions with strofl' fluorescent emissions the total variation of a single determination could approach the 0. level. By using the usual strategy of repeated determinations, any arbitrary precision c be reached.

An XRF sys'tem for on-line analysis of solutions and slurries is diagrammed in Figurt

The system consists of:

1. A high power, rotating target X-ray tube.

2. Monochromator for providing monochromatic exciting radiation.

3. Sample cell connected in a recirculating loop to the solution tank.

4. X-ray spectrometer.

5. Computer for controlling the data taking process.

Each of these components will be considered in detail.

X-Ray Tube

The X-ray tube must generate a well-monitored flux of characteristic radiation such hi after excitation of the sample and the passing of the fluorescent radiation through the t of the X-ray spectrometer, count-rates sufficient to realize the desired analytical precji

within reasonable times are attained. The X-ray power requirement has been estimated frfy

first principles for a typical Savannah River Plant accountability tank, which has a plu

content of 0.4 g/liter—a dilute solution with severe analytical problems.

X LA-6881, Vol. II, E. A. Hakkila et al., "Coordinated Safeguards for Materials Managemei

a Fuel Reprocessing Plant," Sept. 1977.

126

lit

r
An approximate calculation indicates that a generator of 60 kW electrical input should

feet a complete analysis of this dissolver solution in 4 to 8 hours, the bulk of which time

luld be spent in precisely measuring the X-ray emission of the very dilute plutonium. Only

rotating target X-ray tube can absorb such high power input. Rotating target X-ray generors, available only from foreign manufacturers, have proved power capabilities of up to 90

„tl, which is well in excess of the suggested 60 kW. Such machines have been used successfully

i research laboratories for many years. Unfortunately, they utilize obsolete high vacuum

id power regulation technologies, and consequently do not have the long-term reliability and

le X-ray generation stability necessary for use as a dependable, on-line, high-precision

*alytical tool.

A solution to one failure point, the high-speed rotating seal separating target from

:mosphere, has been demonstrated by the use of a Ferrofluidic rotating seal 2

. The use of

lis seal and dry, high vacuum pumps would yield a tube with dependable, long-term performance.

~-ie incorporation of a minicomputer to control and monitor the operating parameters of the

ibe and power supply would permit operation by relatively unskilled technicians. Suitable,

Lgh-wattage power supplies are readily available on the domestic market.

ia r

•5:;

ft

DC

Monochromator

3 The use of monochromatic exciting radiation increases the signal-to-background ratio'

id would substantially reduce the variations resulting from photochemically produced gas

jbbles in the sample. The use of monochromatic radiation also precisely defines the excitaion conditions, and because the theory of interpreting fundamental parameter XRF data is

ased on monochromatic excitation, improvement of analytical accuarcy would be realized.

The monochromator is essentially an X-ray spectrometer of limited angular range. For racticality it must yield maximum flux of selected X-rays, have mechanical and thermal

tability, be convenient to adjust, be well-shielded, and have mechanical features that

acilitate close coupling to the sample cell. The selected beam intensity is critically

ependent upon the diffracting crystal. The best choices for this purpose are the bent

" raphite crystals produced by Union Carbide, which have exceptionally high diffracting powers.

" rovision for accurate monitoring of the monochromatic beam would be included since even the

est regulated X-ray generators exhibit unaccountable intensity variations'*.

Sample Cell

The sample cell would be on-line, and designed to handle samples in solution, as slurried

•"olids 5

, or as mixtures of both. The elemental content of suspended solids, as well as that

if the dissolver solution, would therefore be included in the analysis. The on-line feature

ias three important advantages: samples would not be removed from the recirculating system;

.n analysis continued over a sufficiently long time would effectively, or actually (depending

in tank size) , analyze the entire sample; and the completeness of mixing would be monitored

the XRF analysis. The latter two points are critical for a valid analysis.

The sample, of volume ranging from several tens of milliliters up to the entire contents

)f a holding tank, would be pumped continuously through the cell. Flow-through operation

rould prevent the buildup in the cell of photochemically produced gas bubbles which introduce

malytical errors, and of course, is the only way to implement the use of the analytically

superior large sample. Turbulence in the cell, necessary to ensure homogeneity of slurries,

rould be effected by an internal stirrer or static baffles. The window of the cell must have

resistance to acid, abrasion and radiation, and have maximum transparency to X-rays. The best

window material for X-rays, beryllium, does not have the necessary acid resistance, and would

. 2W. Longley and R. Miller, "A Simple Rotating Anode Generator," Rev. Sci. Instrum. 46:1 (1975).

! 3 Thomas C. Furnas and Robert R. L. Towns, "High Intensity Monochromatic X-Ray Excitation for

Clinical Analysis," Pittsburgh Conference on Analytical Chemistry and Applied Spectroscopy,

I Cleveland, Ohio, March 3-7, 1975.

**Dr. Thomas C. Furnas, personal caveat, 1975.

5 C. R. Hudgens and G. Pish, "X-Ray Emission Analysis of Slurries," Developments in Applied

Spectroscopy 5, 25 (1966).

127
have to be protected. Materials which would serve as protective barriers or as stand-alone B windows must be evaluated. Among these are Mylar and cleaved natural mica, both of which ha I been used as cell windows at Mound with good results.

The irradiation chamber in which the sample cell is contained also serves as the platfc*

for the entrance slit to the spectrometer and as radiation shielding. For highly radioactive

solutions, it must also incorporate sensors for the prompt detection of leaks. Upon detects jj of the leak, fluid flow through the sample chamber would be stopped and the chamber sealed till means of gate valves.

Spectrometer

The spectrometer, a curved crystal (Johsnnson) type for simplicity of optics, must have

robustness, high-setting precision, and shielding against the highly radioactive background

inherent in the materials analyzed. Its design would incorporate high-resolution encoders,

precision stepping motors, and high-precision bearings, resulting in a reliable instrument c advanced capability at reasonable cost. With an analyzing crystal of dispersion and diffrac

tion profile carefully matched to the setting precision of the spectrometer, interf erence-f

i

spectra would be realized.

The performance of the spectrometer is a function of: the analyzing crystal, which is

primarily responsible for resolution and counting efficiency; and setting precision, which

mainly affects counting precision. These two parameters are interdependent. The optimal I values of resolution and setting precision are dictated by the closest lying energies expec

in the X-ray spectrum. The following table (Table I) lists the X-ray lines of interest

(analytical lines underlined), and their 26 setting precision requirement on the spectrometTable I

X-Ray Wavelengths of Elements in Dissolver Solutions

Wavelength 26

Element Transition (h (Degrees)

Y K-al 0.82884 23.75

Cm L-al 0.8289 23.75

Rb K83 0.82921 23.86

Y K-a2 0.83305 23.88

Cm L-a2 0.8411 24.11

Am L-a2 0.860266 24.66

Pu L-al 0.86830 24.90

Sr K-al 0.87526 25.10

Sr K-a2 0.87943 25.22

Np L-a2 0.901045 25.85

U L-al 0.910639 26.13

Th L-al 0.95600 27.46

As K-al 1.17588 33.95

Se K-al 1.10477 31.84

Np Li 1.0428 30.01

Br K-al 1.03974 29.92

u L

t 1.0347 29.77

Based on the requirement that the strontium K-a doublet be completely resolved, the dij*

[ fracting half-widths (W1/2) of commonly used orientations of LiF, silicon, and germanium,

and the 26 setting precisions (A28) required for attainment of 0.1% contribution to the

counting precision for each have been calcuated. These are presented in Table II along wit

the theoretical integral diffracting power (Eo)/I) of each.

For 99% assurance of setting precision, the targeted precision should be about 1/3 A2(jj which for the most stringent case is about 0.0003°. These performance specifications are considerably higher than are found in commercial X-ray spectrometers, which have setting

precisions of, at best, about 0.005°. Commercial X-ray spectrometers, however, are by no Ml means modern. Without exception they use the design philosophy, materials, manufacturing

processes, and angle setting techniques of the immediate post-World War II period. A

spectrometer capable of meeting the high performance specifications of Table II would use ije

128
Table II

T\~\ f f Tart"! Tier Crystal Characteristics

frvsf al On" pnt"3t*i nn Eu)/I

W

1/2

LiF 200 6.68 0.076° U . UU1J

LiF 400 0.67 0.16 0.006

LiF 220 2.35 0.12 0.0046

LiF 420 0.44 0.20 0.0076

Si 111 0.59 0.049 0.0009

Si 220 0.10 0.081 0.0015

Ge 111 0.19 0.047 0.0009

Ge 220 0.083 0.078 0.0015

i ;jst dimensionally stable of high-strength materials and conservatively loaded precision bearigs, but, in particular, it would exploit the highest resolution angular encoders 6 and angle

usitioners under continuous control of a microprocessor.

The X-ray gathering efficiency of the spectrometer is a function of its geometrical efficiency and its relationship to the sample, and to the reflectivity of the analyzing crystal,

aometrical efficiency is determined by the radius of the spectromater , and the area and difraction profile of the crystal. For this study, a sample-spectrometer system with dimensions

2ar those of commercial instruments, using an 8 cm

2 LiF(220) crystal with a reflectivity of

.56%, gives an overall efficiency of 2 x 10 6.

In order to realize the projected system precision, the analyzing crystal must be thermocated. Appropriate temperature control depends upon the crystal and desired counting preision; for the above crystals to contribute 0.1% to the counting precision, the temperature

luctuation would range from 0.1°C to 0.5°C.

The diffracting powers are calculated for ideally imperfect crystals (which give the

ighest integrated intensities) without the attenuating temperature factor. "Ideal imperfecion" in crystals is a hypothetical case which is convenient for calculation of the order of

Electing powers of analyzing crystals. (The diffraction profiles of such crystals would be

npracticably broad in any case.) Crystals with small to moderate imperfections, readily

mailable from domestic sources, show smaller differences of reflecting powers, smaller

Dsolute reflectivities, and narrower diffraction profiles. Since reflectivity is important

n determining the time required for analyses, crystals should be individually selected with

le broadest permissible diffraction profiles for the task at hand.

With a spectrometer of the specified resolution, the only interferences would be incoarent, multiply scattered (Compton) fluorescent X-rays. Incoherent scattering causes a shift

ward longer wavelengths of the scattered X-rays, and has the effect of introducing interring spectra. It is most intense when scattered from light elements, and is not negligible,

slculation of the intensity of incoherent scattering is a complex process that is extremely

ime-consuming on even a medium-sized computer. Fortunately, it can be adequately determined

aforehand for all elements of interest by profiling any X-ray line of wavelength in the

Region of interest.

Process Control Computer

In closed-loop operation, the computer monitors and controls all functions of the instruct. While taking data, it also inspects the data for consistency. In an XRF application,

M*iis would involve a running statistical analysis of the incoming data. Data which did not

onform to predetermined criteria would immediately be retaken. Situations that the computer

^uld not handle, such as equipment malfunction, would be called to the operator's attention

ar correction.

Itek Corp., Newton, Mass.

129
EFFECT OF GAMMA RADIATION ON COUNTING BACKGROUND

The background from gamma radiation with a well-shielded sample handling system can be

expected to be acceptably low. The geometry of the sample chamber will reduce the radiant

flux striking the diffracting crystal by a factor of 4 x 10 . The detector slit, which is out of the direct line of sight to the entrance slit, will pass mainly that radiation incoh

ently scattered from the diffracting crystal—another 4 x 10 * factor. The analyzing cryst

would scatter, at most, about 0.1% of the incident high-energy radiation. Including the

effect of the insensitivity of the detector to high-energy radiation, total attenuation of

gamma radiation reaching the detector would be better than 10 10

. Loading of the pulse pre cessing electronics would thus be negligible, and the pulses, being of high energy, would t rejected by the energy discriminating (PHA) electronics.

XRF ANALYSIS OF PLUTONIUM AND URANIUM IN DISSOLVER SOLUTIONS

In analyzing for plutonium and uranium, the dissolver solution would be pumped contim

ously through the fluorescent sample cell, and the analytical X-ray lines of the internal

standards, plutonium and uranium, would be counted in sequence. Referring to the flow diaj

(Figure 2), steps 1 and 4,, which are background determining steps, are predicated on the

assumption that all elements would be excited by the brehmstrahlung and beta rays of the

fission products. The importance of these steps depends upon the intensity of excitation

the X-ray tube. Either or both may be eliminated if the ratio of internal excitation to e ternal X-ray induced excitation is sufficiently low.

Strontium and yttrium (see Table I) are logical choices for internal standards becaus

their X-ray lines lie near those of plutonium and uranium, and because the addition of the

elements would not perturb subsequent chemical processing. Both strontium and yttrium are present as fission products; hence, the necessity for step 2. Other, less desirable, elem

for internal standards are bromine, selenium, and arsenic.

The availability of several elements for internal standards opens two options: 1) th of more than one internal standard element, which may be necessary to achieve the targeted

analytical accuracy; and 2) separate analyses of the contents of the dissolver and account

ability tanks, using different sets of internal standards for the two samples.

An analysis by the fundamental parameter program of the Naval Research Laboratories (' of the data from step 2 would, by itself, give an approximate analysis of the solution 7

, data of step 5 analyzed by the NRL program with the inclusion of the internal standards ds and corrected for the fission products strontium and yttrium, from step 2, would give the

y final analysis.

ERROR ANALYSIS AND SUMMARY OF PERFORMANCE

The standard deviation of a series of independent XRF analyses is dependent upon: a} the total count accumulated, and b) all other errors. A practical "proper minimization" woulc be one in which categories (a) and (b) contribute equally to the variation. The first is a j rel; mathematical entity. The second category includes mechanical and electronic factors, suclas:

1) counting variations due to setting error (including thermal expansion of components)

;

2) errors in measurement of intensity of the exciting beam; 3) drift in detector electron: s; and 4) variations in the sample —compositional or geometrical. The proper minimization o: each category is a necessary condition fur attaining a desired analytical precision. Eacl of the four factors in category (b) has been assigned a value of 0.1%, the rationale being tit the requisite setting precision is attainable, that electronic equipment matching this pr<l cision is readily available, and that the other variances can be made to attain this prec ion The summed effect of these variables would yield a variation of about 0.20%. This appara .s variation justifies the accumulation of 2 - 2.5 x 10 5 counts—much more would waste timewhich, with appropriate background measurements (approximately equal time counts) , would

effect a precision of about 0.30% for a single determination of a major element, or, becafce

7 J. W. Criss, "HRLXRF, A Fortran Program for X-Ray Fluorescence Analysis," distributed by|

Computer Software Management and Information Center, Suite 112, Barrow Hall, Univ. of Ge pgia Athens, Georgia 30702. -^q
e relatively high contribution of background, 0.4 - 0.5% for a single determination of

y low concentration element. With repeated determinations the variation diminishes inly with the square root of the number of determinations; therefore, time permitting,

sion can be made to reach any arbitrary level.

With the dissolver solution from Savannah River Plant (0.4 g/liter of plutonium with 400

er uranium considered as the dominant absorber) , the described system would yield about

iunts/sec for plutonium. About 80 minutes would thus be required to accumulate 2 x 10 5

s from this very dilute component. For the uranium and the internal standards, the time

sary for the same total count would be 1-2 minutes per element. An estimated 4 to 5

] would be required for a complete single-pass analysis of this dissolver solution.

jFor plutonium at 9 g/liter concentration (dissolver solution of Allied General Nuclear

ices), about three minutes would be required for a 2 x 10 5 count accumulation. Requireof the Atlantic Richland Hanford Company could be met by a system of much more modest

power—about 10% of that needed for Savannah River Plant or Allied General Nuclear.

oth of these cases, the ultimate precision of the XRF system could be reached with

Ite reasonable investment of time.

SAFEGUARDS IMPACT

This XRF system offers an alternative to other analytical techniques presently in use

las several advantages over these techniques. X-ray fluorescence would be performed on- therefore the sampling error would be minimized since the sample size restriction is

nated. On-line analysis would reduce the frequency of transporting highly radioactive

: es to an analytical facility. Reduction in the amount of handling also lessens the risk

mpering with samples. There is no need to remove fission products when using this

[lit, therefore uranium, plutonium and other actinides, if needed, could be determined

1 ut lengthy chemical separations. Undissolved solids could be tolerated in sample

ions, whereas they are intolerable with other techniques. Owing to the high per-

.Lnce components, a considerable improvement in precision over existing XRF systems

be realized with this system.

Concealment of diversion of special nuclear materials would depend on successJ falsifying an analysis; This could, in principle, be done by adding a carefully

. ited excess of standards, which would depress the plutonium assay by a calculated amount.

{> checks between steps 2 and 5 (Figure 2) would impose limits on this strategem. A yet

rigorous cross check would be realized by analyzing for a fission product not readily

.able, in particular technetium, which is present in levels comparable to the fission

icts, strontium and yttrium.

131
Shielding FIGURE 1 - On-Line X-Ray Fluorescence Analyzer.

1. X-rays off: count Pu, U,

Sr and Y

1

2. X-rays on: repeat counts

1

3. add internal standards

1

4. X-rays off: repeat counts

5. X-ray on: repeat counts

FIGURE 2 - XRF Analysis Procedure.

132
Establishing the Traceability of a Uranyl

Nitrate Solution to a Standard Reference Material

by

J. P. CLARK and C. H. JACKSON

Allied-General Nuclear Services, Barnwell, South Carolina 29812

ABSTRACT

A uranyl nitrate solution for use as a Working Calibration

and Test Material (WCTM) was characterized, using a statistically

designed procedure 1 to document traceability to National Bureau of

Standards Reference Material (SRM-960) . A Reference Calibration

and Test Material (RCTM) was prepared from SRM-960 uranium metal

to approximate the acid and uranium concentration of the WCTM. This

solution was used in the characterization procedure. Details of

preparing, handling, and packaging these solutions are covered.

Two outside laboratories, each having measurement expertise

using a different analytical method, were selected to measure both

solutions according to the procedure for characterizing the WCTM.

Two different methods were also used for the in-house characterization work. All analytical results were tested for statistical agreement before the WCTM concentration and limit of error values were

calculated. A concentration value was determined with a relative

limit of error (RLE) of approximately 0.03% which was better than

the target RLE of 0.08%.

The use of this working material eliminates the expense of

using SRMs to fulfill traceability requirements for uranium measurements on this type material. Several years' supply of uranyl nitrate

solution with NBS traceability was produced. The cost of this material was less than 10% of an equal quantity of SRM-960 uranium

metal.

KEYWORDS: Calibration; characterization; reference materials;

traceability standards; and uranyl nitrate solution

INTRODUCTION

Uranyl nitrate solutions are essential for use as standards in the surveillance of

ftical methods, calibration of equipment and methods, and training and testing laborapersonnel. Because of these functions and the importance of uranium measurements for

Jar materials accounting purposes, the standards must be of unquestionable quality and

fly traceable to a primary standard reference material.

Ideally, for a standard to be used as a working calibration and test material it

Id:

1. be stable over extended storage periods,

2. closely match the material routinely analyzed,

3. have an uncertainty consistent with its intended use,

4. be convenient to use, and

5. be economical.

3. Yamamura, et al, NUREG-0253, NRC-13, "Preparation of Working Calibration and Test

arials: Uranyl Nitrate Solution " (1977).

133
Uranyl nitrate solutions, for use as working calibration and test materials, havi

uranium concentrations traceable to the National Bureau of Standards can be prepared i ways. They can be synthesized from well characterized starting materials of known int< such as Standard Reference Materials, using NBS certified mass and volume standards fo ;; solution measurements. Using this procedure, the integrity of the starting material i transferred to the standard, which can be used as either a reference or working calibr,1 test material. The second way uranyl nitrate WCTMs can be prepared is by characterizi

plant uranyl nitrate solution by two different methods of analysis. The NBS traceabil;

is established by calibrating the measurement systems used to characterize the WCTM wi - reference standard while concurrently analyzing both solutions.

The cost of NBS SRMs prohibits their use where large quantities of working standa

required for routine analytical measurement operations. Therefore, the second method

used in preparing a large supply of uranyl nitrate WCTM. The overall cost of procuren

preparation, and characterization was considerably less than the cost of preparing an \ quantity of the standard from SRM-960. By characterizing, the desired reliability was tained and NBS traceability was established.

PREPARATION OF CALIBRATION AND TEST MATERIALS

i Two uranyl nitrate solutions were prepared as calibration and test materials as s in Figure 1. A large quantity of uranyl nitrate solution was prepared for characteriz

and establishing traceability to the standard reference material. This quantity would

the future needs for a working calibration and test material. After the specification

the working standard had been determined, calculations were made to determine how mucV uranium and acid would be needed for the RCTM. Then a small quantity of uranyl nitrat

solution was prepared for use as a reference standard to calibrate the measurement sys used in the characterization of the working standard.

Preparing the Working Calibration and Test Material

Approximately 80 liters of uranyl nitrate were obtained from a plant product tank

used in the preparation of the working standard.

The solution was filtered to remove insolubles and the filtrate was transferred

large drum for mixing. The drum was covered and the solution thoroughly mixed by spai

overnight to ensure homogeneity. Samples were analyzed for concentration of acid, urs and impurities. The results of these analyses were satisfactory. The solution was mi again and systematically transferred to four clean dry 27 liter poly bottles. The net were sealed with a plastic film and closed with a screw cap to protect against evaporj

until the solutions were packaged.

Synthesizing the Reference Calibration and Test Material

The reference calibration and test material was prepared from NBS SRM-960 uraniur

metal to approximate the acid and uranium concentration of the WCTM. The metal was pJ for approximately ten minutes in warm eight molar nitric acid to remove the surface oj

The pickled metal was rinsed with water and acetone and air dried. The required quant

of metal was weighed several times alternately with NBS traceable Class S weights. Tt metal was transferred to a precleaned Erlenmeyer flask fitted with a standard taper ai cooled condenser and reflux head. The metal was dissolved in eight molar nitric acid,

dissolution was completed in 24 hours using low heat. The solution was cooled and quf' tatively transferred to a calibrated bubble neck flask having a screw cap enclosure,

acid concentration was adjusted before the solution was diluted to volume at the desii

temperature. This solution was weighed several times on a double pan analytical balai

using the double transposition weighing technique. The mean weights were determined i corrected to the certified apparent masses of the Class S weights. The makeup value i certainty error was propagated. Table I shows the actual data used and the calculatic

The reference values were also calculated on a volume basis.

PACKAGING THE CALIBRATION AND TEST MATERIALS

The prepared solutions were packaged in various sizes of borosilicate glass ampu!

:ii

liit

liat

Id,

tyle

Fitter

it to

lit

1.!

2.1

c It

i

i

134
ne sealed the day after they were prepared. Specially fabricated 500 ml borosilicate

;s ampuls were used for packaging large quantities of the WCTM for long term storage.

large ampuls will be opened as needed and repackaged in smaller ampuls. The ampuls were

ileaned with hot four molar nitric acid, thoroughly rinsed and allowed to air dry prior

lse. Several 2, 5, 10, 20, and 50 ml ampuls were filled for the characterization work

routine use.

A mechanical ampul filler was used to rapidly aliquot solution to the ampuls, which

: immediately flame sealed with a commercial ampul sealer to prevent evaporation.

:ial precautions were taken when transferring the solution from the large poly bottles

:he ampuls. Care was taken to prevent dripping solution in the ampul necks which could

when the neck was sealed and affect the solution concentration.

The packaging provided 180 large ampuls containing about 13X) grams of uranium which is

valent to approximately five units of SRM-960 uranium metal. Many smaller ampuls were

filled.

PROCEDURE FOR CHARACTERIZING WORKING CALIBRATION TEST MATERIALS

Two analytical chemistry methods known to provide precise and accurate results were

to characterize the WCTM. They were a gravimetric uranium method based on the ignition

ranium to Us0 8 and the New Brunswick Laboratory's modification of the Davies and Gray

imetric method.

Two independent laboratories, selected for their measurement expertise, were contracted

nalyze the solutions. In-house characterization of the two solutions was also performed

g two different analytical methods.

ANALYTICAL MEASUREMENTS

NUREG-0253 lists criteria covering the selection of the number of replicate analyses

characterizing the working standard in section 5.3. The desired limit of error is

ted to the end use of the WCTM and is a function of the accuracy and precision of the

t material to which the WCTM applies. A relative limit of error (RLE) of less than ^

he 0.25% RLE established for the plant material measurements was selected. The number

:amples analyzed determines if satisfactory measurement precisions are attained. In

: acterizing the WCTM by two methods, the limit of error is a function of the precisions

: he two analytical methods used.

A minimum of five replicates is recommended in the NUREG. On the basis of the Safeds Analytical Laboratory Evaluation Program Report of the "Average Within Lab Standard

ation" for the two methods, nine replicate aliquots were selected for the titrimetric

od and six were selected for the gravimetric method. These numbers should have given

oximately equal estimates of the standard deviation of the means, but did not. The

bal precision estimates are shown on page 1 of Appendix I. The in-house measurement

^ision estimates were well known, therefore, six replicate aliquots were chosen

the in-house characterization of the WCTM. An equal number of determinations were pered on both standards by each method.

Evaluation of the Analytical Measurements

, The statistical tests given in NUREG-0253 were applied to the data to compare the conlation analyses and establish an assigned value and limit of error for the working stanThe statistical comparison of the results of the two methods was accomplished folng this scheme:

L. The means and associated standard deviations were calculated for each data set.

2. The precisions of the results obtained by both methods for the WCTM and RCTM were

compared.

3. The two WCTM means were bias corrected for the difference observed between the

RCTM analyses and its makeup value.

4. The WCTM population means, obtained using the two methods, were tested for equality

of means. The approximate variances and degrees of freedom were derived by error

135
propagation.

5. An assigned value and associated standard deviation for the WCTM was calculate

weighting the bias corrected means for both methods.

6. The limit of error and relative limit of error were calculated for the WCTM an tested.

7. The 95% confidence interval was calculated.

The results of the outside laboratories analyses are given in Appendix I with the

tistical treatments listed above. When the WCTM mean results of the two laboratories v, tested using the t-test of equality of means, the difference in population means was si nif icant

.

Investigation revealed that Laboratory II' s analyses of the WCTM had better precis

than the historical precision of the laboratory. Therefore, a pooled standard deviatic was calculated from historical data and characterization data and used instead of the e dard deviation of the WCTM and RCTM results. This was permissible since precision datf

previously obtained for a method used for similar sample analyses can be pooled with

current data to compute a better estimate of the standard deviation with a larger numbe

degrees of freedom.

The results of the outside laboratories appeared to be positively biased from the

up value and in-house analyses of the reference standard. However, no significant difi ence was observed between the laboratories' values for the working standard after corre

for the difference between the observed values and the reference value of the RCTM. Tl results confirmed the in-house assigned value of the WCTM.

SECOND CHARACTERIZATION PROGRAM

The unexplained bias of the reference standard caused in-house concern that oi analytical results may have been negatively biased, so in 1978 a new reference standar(

prepared for a second characterization of the working standard. Table II shows the act data used in calculating the concentration and associated standard deviation values fo: 1978 RCTM.

Several 500 ml ampuls of the WCTM were opened and six aliquots of each standard w< analyzed by two methods as done previously. The analytical measurements and subsequen

statistical treatment of this work are given in Appendix II.

These analyses confirmed the makeup value of the 1978 RCTM and the value previous

assigned to the WCTM. The new assigned value of the WCTM was within the confidence in

val established in the previous work.

Table III summarizes all of the analyses performed in the characterization effort

The mean value of all the analyses of the WCTM is 219.20 ±0.03 mg U/g.

SUMMARY

Detailed instructions are given in the regulatory guide for the preparation and c acterization of a uranyl nitrate WCTM. However, some potential problems that could be

encountered are not specifically addressed. For example, the particular problems that

encountered with Laboratory II 's very precise measurements and the apparent bias betwe

the outside laboratories and in-house results, before they were corrected for, by usinj

techniques described in NUREG-0253.

Following the procedure, a uranyl nitrate WCTM can be prepared from a plant urany

nitrate solution at a minimum cost, with NBS traceability, and within a predetermined

of error.

FIGURES, TABLES AND APPENDICES

Table I. 1976 RCTM makeup value and associated standard deviation calculations.

Table II. 1978 RCTM makeup value and associated standard deviation calculations.

136
ble III. Characterization data summary generated by three different laboratories, using

two different analytical chemistry methods, in two different years, while con- currently analyzing one of two different RCTM's. Each mean has been corrected

for the apparent bias observed in the analysis of the RCTM.

gure 1. Preparation schemes for calibration and test materials.

pendix I. Statistical treatments of the WCTM characterization analyses from outside

laboratories.

pendix II. Statistical treatments of the WCTM characterization analyses by the in-house

laboratory.

137
Table I. 1976 RCTM makeup value and associated standard deviation calculations.

Symbol

A = SA = F =

b = W x = W2 = W 3 = w„ =

s =

A =

Mean Value

Component g/g

assigned makeup value

associated standard deviation

purity of starting material 0.99975

air buoyancy 0.99992

weight of starting material 166.45221

tare weight of the flask 176.695

gross weight of solution & flask 910.209

W3 - W2 = net weight of solution 733.513

standard deviation

(F)(b)(WQ

Standard

Deviation g/g

0.000085

0.0

0.00019

0.034

0.034

0.048

SA^ TT tb " {F "

< SW n

2

) + ( w i)

2 S F2 > + A2 (S„ 2 + SW

2 )] V2

A = (0.99975) (0.99992) (166.45221 g U)

733.512 g

A = 0.226851 g U/g

R 733.512

[(0.99992) 2 {(0.99975)2 (0.00019) 2 + (166.4522)2

(0.000085)2

} + (0.226851)2 {(0.034) 2 + (0.034) 2 }]

SA

~ 2.44 x 10"

Final concentration = 226.851 ±0.024 mg U/g

138
Table II. 1978 RCTM makeup value and associated standard deviation calculations .

Mean Value

Symbol Component

A = assigned value

sA = associated standard deviation

F = purity of NBS-960 metal 0.99975

b = air buoyancy correction 0.99992

Wx = weight of uranium metal 155.37830

W2 = tare weight of flask 164.858 w3 = gross weight of solution & flask 882.044 wu = W3 - W2 = net solution weight 717.186

s = standard deviation

Standard

Deviation g/g

0.000085

0.0

0.00011

0.0083

0.0132

0.0156

a _ (F)(b)(W x )

SA~ §7 & (SW,)

2 + (S Wl ) 2 (S F

2 )} +A2 (SWz

2 + SW3 )]

A = (0.99975) (0.99992) (155.378304 g U) = 0>216578 u/g

717.186 g

SA = n ^, 0 , [(0.99992)2 {(0.99975) 2 (0.00011) 2 + (155.3783) 2 717 . 186

(0.000085) 2

} + (0.216578) 2 {(0.013)2 + (0.00826) 2 }] I*

SA = 1.9 x 10" 5

Final concentration = 216.578 ±0.019 mg U/g

139
Table III. Characterization data summary generated by three different laboratories, using two different analytical chemistry

methods, in two different years, while concurrently analyzing one

of two different RCTM's. Each mean has been corrected for the

apparent bias observed in the analysis of the RCTM.

LABORATORY METHOD CORRECTED MEAN

IN-HOUSE - 76 GRAVIMETRIC 219. 20

OUTSIDE #1 219. 18

IN-HOUSE - 78 219. 24

IN-HOUSE - 76a NBL-MODIFIED D.G. 219. 17 IN-HOUSE - 76b 219. 20

OUTSIDE #2 219. 26

IN-HOUSE - 78 219. 18

THE AVERAGE OF THESE 7 MEANS IS 219.20 MG U/G SOLUTION.

140
WCTM RCTM

COMPUTE WCTM

CONCENTRATION

AND UNCERTAINTY

Figure 1. Preparation schemes for calibration and test materials.

141
APPENDIX I

STATISTICAL TREATMENTS OF THE WCTM CHARACTERIZATION ANALYSES

Data Collected in Confirmation Work From Outside Laboratories

The Mean (M) and Standard Deviation (S) of each set of results were calculated using t; : following equations:

s = (X±

- M) : n-I

n = number of samples in each set

Outside Laboratory I Outside Laboratory II

RCTM WCTM RCTM WCTM

226.99 219.25 226.95 219.39

226.98 219.24 226.99 219.40

227.02 219.38 227.01 219.38

227.00 219.24 226.96 219.37

226.96 219.38 226.98 219.40

227.04 219.32 227.03 219.38

226.98 219.39

226.98 219.39

227.01 219.40

Ma = 226.99 M2 = 219.31 M3 = 226.99 M4 = 219.39

Si = 0.030 S 2 = 0.063 S 3 = 0.025 Su = 0.011

RSDi = 0.013 RSD 2 = 0.029 RSD 3 = 0.011 RSD« = 0.005

F-Test of Precision

Calculated F ratios for both methods are compared to tabulated F values at the 95%

confidence level.

Laboratory I

F = Sij = (0.030) 2 = n .??6 1=4.42

S 2

2 (0.063) z F

FU-cc/2, ni-1, n 2 -l) or F(0.975,5,5) = 7.15 i = 0.14

F Because 0.226 <7.15 and >0.140, the precisions are not different.

Laboratory II

F = -§iLl = (0.025) 2 . 5<6 1=0.178

Su

2 (0.011) 2

F(l-<x/2, nx -l, n2 -l) or F(0.975,8,8) = 4^43 1 = 0.22

Because 5.60 >4.43, the precisions are different.

A pooled standard deviation of .087 was calculated from these and past data for Labor,

142
tory #2 and it was used in all subsequent calculations because the precision for the

KCTM was much better than past measurement performance.

Calculation of Methods Means Based on RCTM

The two WCTM means (X) are bias corrected for the differences observed on the RCTM

analyses

.

The 1976 RCTM reference value (R) = 226.851 mg U/g from Table I

X2 = M2 (R/Mi) = 219.31 (226.851/226.99) = 219.176 mg U/g

X„ = M„ (R/M3 ) = 219.39 (226.851/226.99) = 219.256 mg U/g

= 219.205

Calculation of the Equality of the Means

The approximate variances (V) and degrees of freedom (f) for each mean are calculated.

( s,

2 +

_ UnjMx 2 +

(n

V2 ~ (219.157) :

a)M

(0.030) 2 (0.067

(6)(226.99) z (6)(219.30) 30) 7

0.00088

(n 3 )M3

2 (njM, 2

;

V u ~ (219.256) : (

(0.087) :

(9) (226.99) : 19

(0.087) 2 V

) (219. 39)

2/ 0.0016

Ui Mi } +

\n 2 M2 J n x -l n 2 -l

(0.00088) 2 (219.18)" (O.OSO) 2

^ 2"""

(6) (226.99)

(219. 18)" (0.067) TT2

(6) (219.39)"

s 5 *

/ Xu 2 S 3 "\"

V n 3 M 3 "/ n 3 -l

Xu Su" n„ M4

' nn-1

fu

/ (219.26)" (0.087)

""

\ (9)(226.99)"

(0.0016) :

/ (219.26)" (0.087)'

(9)(219.39)" ~ 8 *

unded to the nearest integer.

143
5 . Compute the T Statistic with f Degrees of Freedom Using the Following Equation:

_ _ X2 - X

J

T = ', =± in. which , . , _ I 219.176 - 219.256] . £n T =

'— 1 = 1.60 + V u JO. 00088 + 0.0016

(V 2 + Vu)^ (0.00088 + 0.0016) 2 f = V 2

Z +

V^ rin which (0.0008) z (0.0016) 7

" - 13 * fa f * 5 8

T from T-Table @ T(l-*/2 t f) = T(0.975,13) = 2.16

Because 1.60 <2.16, the means are not different.

6. Assignment of WCTM Concentration Value (A) by Weighting the Bias Corrected Means fn

each Method

A = W2 X2 + W„X2

1V 2 0.00088

W2 = ! 1 = 1 1 0.6452

v7 +V\ 0.00088 +

0.0016

Wu = 1 - 0.6452 = 0.3548

A = (219.176) (0.6452) + (219 . 256) (0 . 3548) = 219.204 mg U/g solution

7 . Calculation of the Standard Deviation (Sa) Associated with A, with fA Degrees of Fr

1+4 (0.6452) (0.3548)

1761

SA

« 0.027 mg U/g

Na is the number of degress of freedom used in calculating the limit of error for A

1 1 +

W4

Z = (0.64

f 2

z f 4 5 Na z W2

Z +

W u

z = (0.6452) z +

(0.354872

~ ~ 10 *

8. Calculation of the Limit of Error (LE) and Relative Limit of Error (RLE)

LE = 2SA

= 2(0.027) = 0.054 mg U/g

RLE = 100% LE = (0.0540)100% = Q.Q25%

A 219.204

9. Test to Determine if the RLE Meets the Requirement of <l/3 of the Plant RLE of 0.2. which is 0.08%

The target RLE of 0.08% is >0.025% of the WCTM, therefore the WCTM characterizatio:

satisfactory .

* Rounded to the nearest integer.

144
Calculation of the 95% Confidence Interval (CI) for the Assigned Value

CI = A ± t (l-«x/2, Na )(SA)

t (0.975,10) = 2.228

CI = 219.20 ± 2.228 (0.027) = 219.14 to 219.26 mg U/g Solution

145
APPENDIX II

STATISTICAL TREATMENT OF THE WCTM CHARACTERIZATION ANALYSES

1. Data Collected in Confirmation Work from 1978 In-house Analyses

The Mean (M) and Standard Deviation (S) of each set of results were calculated usin

following equations:

. I Xi M =

S = S (Xi - M) 2 n-1

n = number of samples in each set

Gravimetric Results

MiSi

RSDi

RCTM

216.555

216.584

216.509

216.519

216.521

216.658

216.558

0.057

WCTM

219.209

219.153

219.192

219.247

219.324

219.182

219.217

0.061

0.028

Titrimetric Results

RCTM WCTM

216.660 219.057

216.583 219.363

216.490 219.193

216.660 219.057

216.490 219.193

216.607 219.237

M3 = 216.582 M4 = 219.183

S 3 = 0.077 S„ = 0.116

0.026 RSD 2 RSD 3 = 0.036% RSD« = 0.053%

F-Test of Precision

Calculated F ratios for both methods are compared to tabulated F values at the 95%

confidence level.

Method (1)

F . Sjj. (0.057) 2 = Q873

Si 2 (0.061) 2

F(l-<*/2, ni-1, n 2 -l) oi F(0.975,5,5) = 7 . 15 I = 0.14

F

Because 0.873 <7.15 and >0.140, the precisions are not different.

Method (2)

2

I F= 2 Sal. (Q.Q77) . Qj44

S u

2 (0.116) 2 = 2.27

Because 2.27 <7.15 and >0.140, the precisions are not different.

Calculation of Method Means Based on the RCTM

The two WCTM means (X) are bias corrected for the differences observed on the RCTM

analyses

.

146
The RCTM reference value (R) = 216.578 mg U/g from Table II.

X 2 = M2 (R/Mi) = 219.217 (216.578/216.558) = 219.237

Xu = M* (R/M3 ) = 219.183 (216.578/216.582) = 219.179

X2+« = 219.208

Calculation of the Equality of the Means

The approximate variances (V) and degrees of freedom (f) for each mean are calculated.

v2 = x 2

2 f, I

1

2

, +-^-^

V 2 = (219. 237) 2 I (0-057) 2 + (0.061) 2 ) z 0>Q0118

\ (6) (216.558) 2 (6) (219. 217) 2J

S«

2 \(n 3 )M 3

2 (iu)M 2

V, = (219. 179) 2 (0 - 077)2

2 +

(0 ' 116 > 2

? J = 0.00325

\(6)(216.582) 2 (6) (219.183) 2 '

f xv ss fliZIZV

[ iu Mi } + j n2 M2 J rii-1 n 2 -l

(0.00118) 2

/

r (219.237) z (0.057) z V ( (219 237) z (0 . 061) z F

\^ (6)(216.558) z J +V (6) (219.217) z / * 10 *

f« ~

f X4

Z S 3Z AH J* 2

V n 3 M3 l nn Mu j n3 -l n u -l

(0.00325) 2

/" (219. 179)' (0.077) ZT / (219.179) 2 (0 .116)^

I (6)(216.582) z ) + \ (6) (219.183) z ) = 9 *

Compute the T Statistic with f Degrees of Freedom Using the Following Equation:

T = |X 2 - X,

1

in which T = 219.237 - 219.179 = 0 . 871

yv2 + Vu ^0.00118 + 0.00325

(V 2 + VQ 2 (0.00118 + 0.00325) 2 fu ~ V 2

Z

| \'/ in which (0.00118)* ' (0.00325) 2 * 14 *

f 2 f 4 10 9

lounded to the nearest integer.

147
T from T-Table @ t(l-«/2, f) = T(0.975,14) = 2.145

Conclude the two means are not different since 0.871 <2.145

6 . Assignment of WCTM Concentration Value (A) by Weighting the Bias Corrected Means froi I each Method

A = W2 X 2 + W„X„

1_ 1 V 2 0.00118 +

W2 = 1 1 = = 0.73363

V 2 V, 0.00118 0.01

Wii - 1 - 0.73363 = 0.26637

A= (219.237) (0.73363) + (219 . 179) (0 . 26637) = 219.22 mg U/g solution

7 . Calculation of the Standard Deviation (Sa) Associated with A, with fA Degrees of Fre

1

SA ~ W 1 +{4 w2 w, +

W = ^ + |- =1155^5

1+^4 (0.73363) (0.26637) f^jj + ^ h = 0.0318 mg U/g

1155.15

Na is the number of degrees of freedom used in calculating the limit of error for A.

1 1 Na a W2

^ +

Wu

/ ~ (0.73363) r

(0.26637) 2

" = 16 *

f 2 fu 10 +

9

8 . Calculation of the Limit of Error (LE) and Relative Limit of Error (RLE)

LE = 2Sa = 2(0.0318) = 0.0636 mg U/g

RLE = 100% LE = (0.0636)100% = Q Q29%

A 219.22

"

9. Test to Determine if the RLE Meets the Requirement of <l/3 of the Plant RLE of 0.25%

which is 0.08%

The target RLE of 0.08% is >0.03% of the WCTM, therefore the WCTM characterization i satisfactory.

10 . Calculation of the 95% Confidence Interval (CI) for the Assigned Value

CI = A ± t (l-=/2, na ) (SA)

t (0.975,16) = 2.120

CI = 219.22 + 2.12 (0.0318) = 219.15 to 219.29 mg U/g solution

* Rounded to the nearest integer.

148
Preparation of Uranium Standard Solutions for X-Ray Fluorescence Analysis*

by

C. M. Wong, J. L. Cate and W. L. Pickles

Lawrence Livermore Laboratory, Livermore, California 9^550

ABSTRACT

A method has been developed for gravimetrically preparing uranium

nitrate standards with an estimated mean error of 0.1% (1 sigma) and a maximum error of 0.2% (1 sigma) for the total uranium weight.

Two source materials, depleted uranium dioxide powder and NBS Standard

Reference Material 960 uranium metal, ' were used to prepare stock

solutions. The NBS metal proved to be superior because of the small but

inherent uncertainty in the stoichiometry of the uranium oxide. These

solutions were used to prepare standards in a freeze-dried configuration

suitable for x-ray fluorescence analysis.^ Both gravimetric and

f reeze-drying techniques will be presented. Volumetric preparation was found to be unsatisfactory for 0.1% precision for the sample size of

interest.

One of the primary considerations in preparing uranium standards for

x-ray fluorescence analysis is the development of a technique for dispensing

a 50-ul aliquot of a standard solution with a precision of 0.1% and an accuracy of 0.1%. The method developed corrects for variation in aliquoting

and for evaporation loss during weighing.

Two sets, each containing 50 standards have been produced. One set has

been retained by LLL and one set retained by the Savannah River Project

(SRP).

KEYWORDS: X-Ray fluorescence analysis; actinides; high accuracy; freeze

dried; particle size effects: gravimetric standards; radioactive

samples

INTRODUCTION

The purpose of this study was to develop a method of preparing uranium standards for

ay fluorescence analysis to meet the requirement of the Safeguard Programs at Lawrence

.ermore Laboratory (LLL). Some of these requirements were that the standards source

erial must be of high purity; the method of preparation of the standard solution must be

:urate; the aliquoting technique must be precise; the standard configuration must be

table for x-ray fluorescence measurement; and, the maximum error associated with all

ects of preparation must be minimal and well known.

This work was performed under the auspices of the U.S. Department of Energy by the

rrence Livermore Laboratory under contract No. W-7405-Eng-48.

.s suggested by S. S. Yamamura, private communication.

'. L. Pickles and J. L. Cate, Jr. Quantitative Nondispersive X-Ray Fluorescence Analysis

Highly Radioactive Samples for Uranium and Plutonium Concentration, Lawrence Livermore

oratory, Report UCRL-47417 (1973) , and in Advances in X-Ray Analysis, C. L. Grant, C. Barrett, J. B. Newkirk, and Co., 0. Rudd, Eds. (Plenum Press, New York, 1974), Vol. 17,

337-347.

149
I. STANDARDS SOLUTIONS PREPARATION

A. Uranium Oxide Standards

To make our initial trial uranium standards, we used some 15-year-old depleted urai

oxide that was 99.93% pure UO^. The sample was spectrographically analyzed for

impurities (see Table I).

The uranium oxide was initially prepared according to the following standard method

(1) The oxides are converted to U3O8 by heating in a cleaned platinum crucible at

800 C for one hour; (2) The crucible is removed from the oven and the powder stirred ar mixed well; (3) The crucible is returned to the oven and heated for one hour; and (4) U3O3 is allowed to cool in a desiccator and then weighed. The weight of the

UO2 before heating was 201.859 g and the weight of U3O8 after heating was

208.967 g. There was a weight gain of 7.108 g which indicates a 99.6% theoretical

conversion of UO2 to U3O8, and according to the equation 3 UO2 + O2 *

U3O3, the gain in weight should have been 7.973 g. This difference may indicate

that the original UO2 contained a small amount of U3O3 . The stoichiometry as well as the isotopic composition were established by mass spectroscopy.

A weighed aliquot of the U3O3 was dissolved in sub-boiled distilled 8 N HNO3, and a stock solution was prepared using sub-boiled distilled H2O to dilute to volume at 1 N HNO3 . Sub-standards from 0.1-10 mg U/g of solution were then prepared

from the stock solution.

All aliquots and dilutions were made by powder weight. Aluminum foil pans and funr were used for initially weighing the UO2 and U3O3 to eliminate the static

charge that the powder exhibited. Next, the solutions were pipetted, weighed,

f reeze-dr ied , and analyzed by x-ray fluorescence.

Table I. Spectrochemical Analysis for

Impurities in UO (ppm)

.

Fe 100 Al 5 B <1 Ni 30 Mg 5 Sr <1 Ca 20 Cu 4 Pb 0.5

Mo 20 Mn 3 Ti <0.3

Cr 15 Sn 3 Be <0.1

Si 15 Zn 2 V 1 The following elements were not detected:

Detection limit

Ba <30

Zr <10

Nb, Sb <3 ! Bi, Co <1 Cd <0.3 'j

150

I. f
B. Uranium Metal Standards

The uranium metal, NBS Standard Reference Material 960, was obtained from the National

'eau of Standards and cleaned of surface oxide according to the following NBS procedure:

i dip the uranium metal in 1 : 1 HNO3 for ten minutes to remove surface oxide; (2)

#se in distilled water; (3) etch in 1:3 HC1 for five minutes; (4) rinse thoroughly in

stilled water; and (5) remove excess distilled water and place in a vacuum desiccator to

jain constant weight. Removal of surface moisture is accelerated and reoxidation of the

;al surface is retarded by drying under vacuum for a suitable length of time

oproximately 1/2 hour).

After constant weight is attained, a weighed aliquot of the uranium metal was

solved in sub-boiled distilled 8 N HNO3 and sub-standards were prepared. These

.utions were pipetted, weighed, f reeze-dried, and analyzed by x-ray fluorescence in the

le manner as the uranium oxide. The metal-based standards proved to be far more precise

in the oxide-based standards.

3

II. WEIGHING ACCURACY

We checked our H-16 Mettler balance and our weighing method to determine the

isistency of the balance in weighing and the individual bias in reading the weights.

f*st, one person alternately weighed a series of empty vials, and then a series of vials

:h containing 50 ul of distilled water. This procedure was repeated ten times for each

,il. Second, two people did the weighing, rezeroing the balance and alternately reading

; weights. The results in Table II show that the slight variance in weight read by two

>ple or by the same person is not great enough to exceed the 0.1$ precision and accuracy

weight that we hope to achieve.

Table II. Average Deviation in Reading Weights

Reading by Empty Vials Vials + H

20

One person ± 0.00002 g ± 0.00003 g

Two people ± 0.00003 g ± 0.00005 g

III. DISPENSING PRECISION AND ACCURACY

The accuracy and precision of three high-precision microdispensers were measured by

ghing each aliquot on a Mettler H-16 balance. To standardize the effects of weight loss

>m evaporation while weighing, the following procedure was used: (1) open vial; (2)

rt electric timer; (3) pipet 50^ul aliquot of distilled water; (4) dispense sample

.0 vial; (5) place on balance pan; (6) weigh; (7) read weight when timer is exactly

i0 min. The Brand A pipet proved to be far superior to the Brand B and Brand C pipets,

shown in Table III. However, none of the three pipets are capable of delivering 0.1%

:uracy or precision volumetrically , therefore, the accuracy had to be achieved

ivimetrically , weighing each sample individually.

r . L. Pickles, J. W. McClure, R. H. Howell, A Non-Linear Method for Including the Mass

ertainty of Gravimetric Standards and the System Measurement Errors in the Fitting of

ibration Curves , Lawrence Livermore Laboratory Report, UCRL-80375 (1977), and in the

'ceedings of the American Chemical Society, Symposium on Non-Destructive and Analytical

^mical Techniques in Nuclear Safeguards (1978), 175 ACS.

151
Table III. Pipet Precision and Accuracy for Delivery of 50 pi Water Samples

Number of % Accuracy3 Average % Pipet Measurement Range Av Precision

Brand A 70 -1 .38 to 1.72 -0.48 0.47

Brand B 60 -6 62 to 7-28 -2.27 3.03

Brand C 20 -9 46 to 1.16 -1 .42 2.65

a o Corrected for H^O density variation at room temperature (22 C),

IV. EVAPORATION CORRECTION IN WEIGHING

The evaporation rate of 50^ul aliquots of several solutions was determined by

repeated weighings over a five-minute interval for each sample. It is apparent from Fig

that the evaporation loss becomes significant for a weighing time greater than 15 s.

52

51

£

co

o° 50

a. o

49

48

19.2 mg U/g

a

o

°o°o

°

O O

O O

0 1 2 3 4 5

0.0988 mg U/g

O O

O O

O O

Of H 2°

0 1 2 3 4 5 Time after dispensing — min

Fig. 1. Weight changes of solution during weighing.

152
The evaporation rates of 50 ul of pure water and of 100 ppm, and 19000 ppm (in 0.25

N HNO3) uranium solutions vary between 0.19 and 0.21 mg/min (a slightly lower rate of

0.15 mg/min was observed for 6 N HNO3). In actual practice, the minimum time required

for conveniently dispensing (with an automatic pipet) and weighing the 50-ul aliquot is

about 30 s. Therefore, within 30 s, 0.10 mg of solution would have been volatilized, or

0.2% of a 50-mg aliquot, for solutions with the above evaporation rate. To achieve the

0.1$ accuracy, as desired by the Safeguards Programs, it is necessary that this kind of

correction be incorporated into the weighing procedure.

The 50-ul aliquot was chosen because this volume best meets the requirements of the

sample- mounting and counting procedure developed by LLL for x-ray fluorescence analysis of

uranium-bearing solutions. 2 Briefly, this procedure involves mounting an adhesive

Kapton polyimide film on a 35-mm slide with a 1.27-cm diameter circle of polycarbonate film

centered on the adhesive side of the Kapton. The polycarbonate film was wiped with a dilute (1:100) solution of aerosol-OT-dispersant, then dried and tared. The standard

solutions were then dispensed onto the circle of film, weighed continuously every 30 s for

five minutes, freeze-dried, and covered with another layer of adhesive Kapton (Fig. 2).

After the procedure was completed the samples were analyzed by x-ray fluorescence.

V. FREEZE-DRYING TECHNIQUE AND PHYSICAL SAMPLE CHARACTERISTICS

After weighing, the sample was quick-frozen by placing it on an aluminum chill block

that was at LN temperature. The chill block and sample were then placed in the vacuum

apparatus and kept under vacuum until the sample was completely dry — from 6 to 16 hours

(Fig. 3).

35 mm slide front frame

Adhesive coated polyimide cover

Sample

Polycarbonate

Adhesive coated polyimide base

35 mm slide backing frame

Fig. 2. Mounting procedure for an x-ray fluorescence analysis

sample. This is a convenient

procedure that provides primary

containment for hot samples.

Fig. 3. Freeze-drying apparatus.

The apparatus consists of a vacuum

manifold, a sample evacuation jar,

a foam insert, and an aluminum

heat sink at LN temperature.

153
Some of the sample standards were gold coated by vacuum-deposition techniques and th

crystals examined under various magnifications with a mini-scanning electron microscope.

The optical photographs showed details of the fibrous, snow-like crystalline structure

which is an important consideration for understanding the type of particle-size effects t< expect in the x-ray fluorescence analysis of these samples (Fig. 4). The diameter of the

sample is 12.7 mm. Figure 5 is a magnified view of the edge of the sample showing the

fibrous nature of freeze-dried U-NO3.

The use of excessive concentrations of dispersant in preparing these samples was discontinued after discovering that it caused the center portions of the sample to freeze

in an irregular way (Fig. 6).

Fig. 4. Optical photograph of a 12.7-mm-diameter freeze-dried

U-NO sample.

Fig. 5. Highly magnified, optical

photograph of a freeze-dried

U-NO3 sample. Note its

fibrous nature.

Fig. 6. Freeze-dried U-NOo

sample that has been collapsed to expose the interior structure.

154
Most of the information about the structure of the freeze-dried samples was obtained

by physically crushing the center and photographing the exposed interior using high

magnification. As shown in the fissure on the left in Fig. 7, freeze drying U-NO3

results in long twisted fibers that are generally perpendicular to the plane of the

sample. A simple exponential absorption model, with an increased effective sample

thickness, may be used to describe the effect of freeze-dried U-NO on the XRFA

instrument response.

3

SUMMARY

In conclusion, we have demonstrated that highly accurate, freeze-dried U-NO3

standards that are suitable for x-ray fluorescence analysis can be prepared in the mass

range from 4 yg to 1000 ug. Gravimetric methods, using NBS 96O U metal, have proved to

be the most accurate preparation technique; standard reference material accuracies of

approximately 0.2? can also be achieved by this method.

NOTICE

"This report was prepared as an account of work

sponsored by the United States Government.

Neither the United States nor the United States Department of Energy, nor any of their employees, nor any of their contractors, subcontractors, or

their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness or

usefulness of any information, apparatus, product

or process disclosed, or represents that its use would not infringe privately-owned rights."

155
Use of a Non-Linear Method for Including the Mass Uncertainty of Gravimetric

Standards and System Measurement Errors in the Fitting of Calibration

Curves for XRFA with Freeze-Dried UNO3 Standards

W. L. PICKLES, J. W. McCLURE and R. H. HOWELL

Lawrence Livermore Laboratory, Livermore, California 94550

ABSTRACT

At LLL we have used a sophisticated non-linear multiparameter

fitting program to produce a best fit calibration curve for the

response of an x-ray fluorescence analyzer to uranium nitrate, freeze

dried, 0.2% accurate, gravimetric standards. The program is based

on unconstrained minimization subroutine, VA02A. The program con- siders the mass values of the gravimetric standards as parameters to be fit along with the normal calibration curve parameters. The fitting

procedure weights with the system errors and the mass errors in a consistent way. The resulting best fit calibration curve parameters

reflect the fact that the masses of the standard samples are measured

quantities with a known error. Error estimates for the calibration

curve parameters can be obtained from the curvature of the "Chi-Squared

Matrix" or from error relaxation techniques. We have shown that non- dispersive XRFA of 0.1 to 1 mg freeze-dried UNO3 can have an accuracy

of 0.2% in 1000 sec.

KEYWORDS: Non-linear least squares, x-ray fluorescence

analysis, uranium, two-dimensional calibration.

INTRODUCTION

The goal of the work described here was to be able to produce highly accurate (0.1 to 0.2%) calibration curves of non-destructive assay instruments where the accuracy of the

standards available is the limiting factor, or at least a major source of calibration error.

In reaching the ultimate accuracies possible for a particular NDA measurement system,

the instrument long-term precision is often not the limiting factor. The variability of

sample preparation and the accuracy and applicability of the standards used for calibration

of the instrument usually create the greatest source of uncertainty.

We have developed a mathematical method of dealing with these types of errors in a statistically correct way. Our first test of this method was with standards accuracy for

x-ray fluorescence analysis of f reeze-dried^ UNO3. The method can also be used to evaluate

the importance of sample variability errors. The type of computer code we have used in

this method is commercially available from several sources^ as a package which requires

only a small amount of input-output user generated software.

METHOD

Our LLL XRFA system^ has a repeatable precision which has been measured to be 0.1%

(two standard deviations). In attempting to utilize this system for accountability measure- ments in the nuclear fuel cycle, we were continually frustrated by the lack of high accuracy

solid samples in the mass range from 10 to 1000 yg. We were finally able to produce UNO3

standards by a freeze-drying method with an NBS traceable accuracy of 0.2% (one standard

deviation).^ These samples were thought to have particle size absorption, but because of

the uniform fibrous nature^ of these freeze-dried samples it was expected that these

absorption effects would be calculable to high accuracy. We have used 100 of these

standards to calibrate our XRFA instrument.

Since the mass accuracy error of the standards was estimated to be twice as large as the instrument precision errors, we felt it was particularly important to include the mass

156
icertainty in the calibration procedure. Our approach was to treat the mass values of

te standards in exactly the same way as we normally treat the instrument's response to

iose standards. That is, the mass value of each standard is a gravimetrically measured

.antity. The gravimetrical mass value, Mi, is not the "true mass" of the standard. It

.ffers from the true mass in a normal way. The gravimetrically measured mass, M-^, has a

% probability of deviating from the true mass value by less than 0.2%.

eated a set of parameters which represent the true mass values,

We therefore

.ere is one X-^, or true mass, for each standard. It is now possible to use these new para-

:ters in expressing the instrument response calibration curve equation, YFUN.

YFUN = G(A,B,C,y

1 ,y 2 ,X

i

)

e true mass X-^ of the standard is one of the variables in the calibration function instead

being a fixed constant. Consequently, the true mass,

d ]S2 > the "usual" calibration curve fit parameters.

may be fit along with A,B,C,yi

The result of this technique is to start from a set of gravimetrically measured standard

ss values and measured XRFA instrument responses to those standards and arrive at both the

st probable, or true mass, of the standard, and the most probable system response values,

lis is diagrammed schematically in Figure 1.

Most probable

standard value

YFUN

YFUN

Measured system

response value

Most probable system

response value

Measured

standard

value

Figure 1. Overall result of non-linear least squares fitting is a most probable system response value and a most probable

standard mass value.

The fitting procedure is accomplished by a commercially available, 2 non-linear, unconrained minimization, computer program. The program minimizes the quantity chi-squared.

r chi-squared not only involves the deviations in the instrument response from the caliation curve as is usual, but must also include the deviations of the gravimetric mass

lues from the true mass. The value of chi-squared per degree of freedom is a measure of

e "goodness" of fit of the calibration curve and true masses to all the experimental data,

r chi-squared is defined in Figure 2. The expression for chi-squared has two sums of

ighted, squared deviations. The first of these terms is similar to the whole expression

ied for fitting programs in the past. It is different in that the true mass, X-^ is used

place of the gravimetrically measured mass, M^. The second term is new, and is the sum

157
TO CONSISTENTLY USE BOTH TYPES OF ERRORS

CHISQUARED MUST INCLUDE THE STANDARDS

MASS ERRORS L3

System response

R. — YFUN (A r

B

2

, Ml ,M2 ,C,

2 0-.

2 Almost normal

CHISQUARED

Standards variations

2e

New contribution

to CHISQUARED

Figure 2. New two-dimensional definition of chi-squared used in the

non-linear fitting technique. Note the use of true rather

than gravimetric mass.

of the squares of the deviations of the measured mass from the true mass, weighted by the

gravimetric errors. The actual calibration curve function, YFUN, which we used in this won!

is shown in Figure 3. The function contains three terms; the first term is a constant, the't! second is a term that represents simple mass absorption, and the third term allows for

absorption in the long thin fibers of UNO3 oriented perpendicular to the plane of the sampl'i

The fact that the free parameters in this function A,B, C,\i\,\i2 , {Xj_} appear as products and 1 that the expression for chi-squared contains X^'s which are free parameters, dictates the use of a non-linear fitting program.

FOR XRFA OF FREEZE DRIED UNO STANDARDS WE CHOOSE A

PHYSICALLY REALISTIC MODEL IS

— Fibrous nature of standard

creates particle size effects

ExciterYFUN = A + B[1-e-'J

i m

]+ C[1-e"^2m

]

^Background ^Normal mass

absorption • Thus the function becomes

YFUN = A + B[1-e"^i x i] + C[1-e-^2 *i]

brous particle

size absorption

Figure 3. The actual calibration curve used in this work has a term for

normal mass absorption and also a term for fibrous particle

size absorption.

158
RESULTS

The final results of using this technique is a "best-fit" value for A,B,C,yi,U2

1 all the Xj/s as shown in Figure 4 numerically and Figure 5 graphically. As can be seen

Figure 5, 40% of the under response is due to simple mass absorption and 60% is due to

rticle size absorption.

THE FINAL RESULT OF THIS TECHNIQUE IS

• A set of most probable model parameters

YFUN = 1.02 X 106 + 4.6 X 105 [1-e" 21 * 10 ~ 4 X] + 5.6 X 105 [1-e- 73x 10 ~ 4 X]

I Background Normal

mass absorbtion

Fibrous particle

size absorbtion

• A set of most probable standard values

X DATA

4 . "54E+00

<•-'

. 459E+00

9 . 521 E-< 00

X FIT

4 . 294E+00

4 . 341 E +00

9 . 594E+00

RES/SIGMA

1 599E+00

1 . 1 86E+00

-7 . 339E-01

•

SIGMA

1 . 000E-01

1 . 000E-01

1 . 0O0E-01

•

Figure 4. Actual numerical results of our fitting method showing our best fit parameters.

PARTICLE SIZE AND MASS ABSORBTION ARE APPROXIMATELY EQUAL LH

Mass of standard

Figure 5. Graphical representation of our best fit calibration curve showing the normal mass absorption, 40%, and the fibrous

particle size absorption, 60%, under responses from the

linear.

159
ERRORS

A non-linear least squares fitting program does not use simple matrix inversion to obtain a unique best fit value for each free parameter and consequently does not produce aj unique error matrix for the free parameters. However, estimates of the overall error is

possible by two methods. In the first method, the curvature of chi-squared space near the best fit value of each parameter is an indication of the sensitivity of the fit to that

parameter. The second and more useful method is to relax the errors on the gravimetric

masses and/or the instrument response precision until a chi-squared per degree of freedom

of approximately three is obtained. A chi-squared per degree of freedom of three means th probability that all the fit parameters are within one standard deviation of their "correc value is 67%. We were able to obtain a chi-squared per degree of freedom of three by re- laxing the instrument response errors to 0.2%. The conclusion we draw from this is we should accumulate counts on an unknown sample until the precision of the response is 0.1%

and then the error we assign to the measurement of that sample will be 0.2% (1 sigma)

.

I SUMMARY

We have found the non-linear fitting techniques as described here to be a powerful

method of creating realistic calibration curves for an NDA instrument and a particular

standards set. The method uses both the gravimetric mass errors and the instrument respon errors in a statistically consistent way. It incorporates the independent gravimetric

measurement of the standards in the calibration curve parameters thus extracting all the

experimental information available for the instrument response and the standards set. It determines the actual most probable value of each standard mass. It allows sensitive

selection among the calibration curve models. It eliminates the need to cross measure

standards, and it allows a realistic appraisal of the overall accuracy error of an NDA

instrument and it's standards.

REFERENCES

1. C. E. Pietri, J. S. Puller and C. D. Bingham, "The Chemical and Isotopic Analysis of

Uranium, Plutonium, and Thorium in Nuclear Fuels," Proceedings of the ANS Topical

Conference on Analytical Methods for Safeguard and Accountability Measurement of

Special Nuclear Material, 1978.

2. C. M. Wong, J. L. Cate, and W. L. Pickles, "Preparation of Uranium Standards for X-Ray

Fluorescence Analysis," Proceedings of the ANS Topical Conference on Analytical Method

for Safeguard and Accountability Measurement of Special Nuclear Material, 1978.

3. International Mathematic and Science Library—XCSSQ Routine.

4. National Bureau of Economic Research—NL2S0L Routine.

5. W. L. Pickles and J. L. Cate, "Quantitative Nondispersive X-Ray Fluorescence Analysis

of Highly Radioactive Samples for Uranium and Plutonium Concentration," Advances in

X-Ray Analysis Vol. 17, Plenum Publishing Corp., New York, 1973.

*Work performed under the auspices of the U.S. Department of Energy for the Lawrence

Livermore Laboratory under contract number W-7405-ENG-48

.

NOTICE

: :

S6\

"This report was prepared as an account of work

sponsored by the United States Government. Reference to a company or product

Neither the United States nor the United States , , , Department of Energy, nor any of their employees, nam e aoes not imply approval or

nor any of their contractors, subcontractors, or recommendation of the product by

their employees, makes any warranty, express or , , , .. c „ ... . implied, or assumes any legal liability or respon- tne University Ot Lalllornia or the

sibility for the accuracy, completeness or U.S. Department Oi Energy to the

usefulness of any information, apparatus, product . . . u

or process disclosed, or represents that its use exclusion Ol Others that may be

would not infringe privately-owned rights." Suitable.

160

tin
Future Approaches to Material Control and Accounting

by

T. S. SHERR, G. D. SMITH, and L. F. WIRFS

Division of Safeguards, U.S. Nuclear Regulatory Commission

Washington, D.C. 20555

ABSTRACT

I This paper presents a short description of the safeguards responsibilities and

activities of the U.S. Nuclear Regulatory Commission (NRC), the NRC regulatory

requirements for safeguards in the area of material control and accounting

(MC&A), and the current NRC efforts which may result in significant changes in

the current U.S. safeguards system. The preliminary results of NRC staff and

contractor MC&A activities will be discussed, as well as the recommendations

of a recent NRC task force on MC&A.

CRDS: Nuclear safeguards, material control and accounting, fuel-cycle regulation

INTRODUCTION

U.S. Nuclear Regulatory Commission is responsible for ensuring that there is appropriate

lection against radiological sabotage at commercial nuclear facilities and theft of ear materials. The Commission meets this responsibility through a safeguards program

rised of four major elements —regulatory development, licensing, inspection and evaluaof licensee operations, and coordination with other agencies. Material control and

unting activities comprise a significant part of the NRC program in each of these

ents. As background prior to addressing future approaches to material control and

unting in this paper, a brief overview is first provided of the current NRC safeguards

ram and activities.

first major element of the NRC safeguards program is the development of a regulatory

:. This element includes developing safeguards policies, requirements which reflect

e policies, and guidance to assist licensees in meeting these requirements. In this

, several major policy activities and supporting technical studies are underway which

ess MC&A. These studies will be described in detail later in this paper.

second major element of the NRC safeguards program is licensing which includes the

ew and approval of proposed safeguards plans in order to assure that appropriate safeids will be implemented. Reviews of such plans occur when licensees respond to new lations, when applicants apply for new licenses, or when application is made for license

fication or renewal. In this area, the major current licensing activity for MC&A is the

ementation of new requirements for measurement quality control, which includes specific

bration and statistical procedures.

third major element of NRC's safeguards program is the inspection and evaluation of the

guards systems, procedures and practices of current licensees. These NRC activities

two objectives: (1) to insure that current safeguards practices are in compliance with

requirements, and (2) to insure that NRC's requirements themselves are adequate through

evaluation of the vulnerabilities and capabilities of present systems. In this area,

is currently conducting a major evaluation program at facilities possessing plutonium

high-enriched uranium (greater than 20% in the U-235 isotope). Included in this program

comprehensive evaluations of the material accounting and internal control systems at

e facilities. The evaluation findings and recommendations are fed back into the reguladevelopment and licensing activities previously discussed.

161
The final major element of the NRC safeguards program encompasses the activities which NRC accomplishes with extensive coordination with other government agencies. These activities

include threat assessment, planning for nuclear emergencies, and international safeguards

matters. In the international area, NRC is currently preparing a new regulation to implement the US/IAEA safeguards agreement which will provide for IAEA inspection at selected

facilities and will require that certain information be reported to the IAEA. The information to be reported includes design information and data concerning inventories, inventory

changes, and material balance reports, among others. Certain U.S. facilities are already

supplying design information, on a voluntary basis. Once the agreement comes into force,

all eligible facilities will be required to report design information and material account- data .

SUMMARY OF CURRENT REGULATIONS

All of the current NRC regulations concerning material control and accounting for special

nuclear material are contained in Title 10 of the Code of Federal Regulations under Part 7( It is contemplated that the requirements needed to implement the pending US/IAEA agreement

for safeguards will be published in Part 75.

Most persons and organizations authorized to possess at any one time and location more thai one effective kilogram* of special nuclear material in unsealed form are required to comply

with comprehensive material control and accounting regulations. An applicant must include

in his license application a full description of his proposed program plan to show how

compliance with the regulations will be accomplished. After review and approval by the NRI the plan is incorporated as a condition of the license.

In discussing the control and accounting of SNM, it should be recognized that the first lii of protection against theft or diversion is provided by a comprehensive physical security

program. Classical security measures are utilized, including perimeter barriers and alarm

armed guards, access controls, vaults, portal monitors, surveillance and secured communications. Time does not permit a detailed discussion of physical security programs in this

paper.

Material Control

The material control measures used for safeguarding SNM are similar in many respects to those used by industry for the internal control of valuable assets. The basic features of

the present NRC material control program are summarized as follows:

Assignment of overall responsibility for the material control and accounting program

a single individual

Separation of staff functions to provide internal checks

Approval and maintenance of a manual which specifies current safeguards procedures

Designation of specific control areas for the physical and administrative control of

material in sufficient number to localize losses

Assignment of custodial responsibilities for SNM within each control area to single

designated individuals

i Establishment of storage and handling controls to provide current knowledge of the

identity, quantity and location of all SNM in discrete items and containers

*"Effective kilograms of special nuclear material" means: (1) For plutonium and uranium-2

their weight in kilograms; (2) For uranium with an enrichment in the isotope U-235 of 0.0

(1%) and above, its element weight in kilograms multiplied by the square of its enrichmen

expressed as a decimal weight fraction; and (3) For uranium with an enrichment in the

isotope U-235 below 0.01 (1%), by its element weight in kilograms multiplied by 0.0001.

162
Unique identification of items

Annual audit of the material control program

Material Accounting

ccounting program for special nuclear materials contains many of the basic elements of

ntional accounting applied in the business community. Special emphasis is placed on easurement of SNM for generating accounting source data. The basic features of the

nting program may be characterized as follows:

Physical measurement of SNM received, produced, transferred between control areas,

shipped or discarded

.Measurement quality control program to assure the quality of accounting data and to generate measurement uncertainty information

Material balance and item control areas to assess and localize losses through the

accounting system

Documented system for transfer of material between control areas

.Physical inventories based on measurements

Centralized accounting system employing double-entry bookkeeping

Material balance closures with accounting adjustments to incorporate physical inventory

results

i Statistical limits based upon measurement errors

Evaluation and investigation of differences between the material records and the

physical inventory

Review and reconciliation of statistically significant shipper- receiver differences

'Immediate investigation of losses of discrete items and containers

Annual accounting and measurement system audits

[ Retention of records for five years

licensees are presently required to establish and maintain plant accountability systems

that the measurement uncertainties (limits of errors) associated with a plant material

ce do not exceed 0.5% of material throughput.

cal inventories and material balance closures must usually be performed bimonthly for

>nium and high-enriched uranium and semiannually for low enriched uranium.

Measurement Quality Control

, prehensive quality control program is required by current regulations on physical

( rements utilized in the safeguards accounting program. Weight, volume and nondestrucassay measurements performed on the production line and on shipments and receipts for

intability are covered by these requirements as well as analytical laboratory measure-

. The requirements are summarized as follows:

| Assignment of overall responsibility for the program to a single individual

Performance of process and engineering tests to assure adequate mixing and sampling

Engineering analyses and evaluations to assure adequacy of each measurement system

163
Ongoing use of standards for calibration and control of systematic errors

Generation of current data for estimation of random and systematic errors for each

material balance period, and for the establishment of bias corrections

Statistical control, including control charts, over each measurement system

Investigation and correction of performance outside of control limits

Discontinuing measurements not covered by the quality control program for SNM accou

Retention of records for five years

Maintenance of training and personnel qualification programs

Annual management reviews and system audits

CURRENT DEVELOPMENT PROGRAMS*

MRC's current development programs for material control and accounting can be separated

the following four categories: (1) development of improved evaluative methodologies, (2

development of new approaches to performing material control and accounting operations,

development of new techniques for analyzing material control and accounting data, and (4 development of improved measurement and measurement quality control techniques. Each of

these four categories of the current development program will be summarized in this sect'

Evaluation Techniques

Three evaluation techniques are currently being assessed by the NRC: (1) Diversion Path

Analysis (under development for the Department of Energy by the National Bureau of

Standards), (2) the Material Control System Assessment Procedure (under development for

by Lawrence Livermore Laboratory), and (3) Insider Collusion Methodology (under developrm

for NRC by Science Applications, Inc.).

Diversion Path Analysis (DPA)

DPA is a structured procedure developed by the National Bureau of Standards for the Depar

ment of Energy to systematically consider methods (paths) by which special nuclear materi

(SNM) can be covertly removed from its authorized location in a facility (Ref. 2). It considers various methods by which the removal could be concealed and identifies operatic

procedures or safeguards mechanisms necessary to detect such diversion attempts. DPA dei

predominantly with those insiders who have hands-on access to material and/or records, m

also directs consideration to other personnel who may be in the area. By forcing a detai

and systematic, although judgmental, examination of the facility, assurance is gained tha all diversion paths within the scope of the analysis have been considered.

Materia l Co ntrol System Ass e ssment Procedure

Development of a systematized means to determine the effectiveness of material control

systems is the major thrust of this work sponsored at the Lawrence Livermore Laboratory

(LLL) by the NRC (Ref. 3). It is LLL's intent to provide the needed assessment tools and

procedures, and the data base necessary for application of the assessment methodology. ,

The methodology is intended to consider the integration of material control, material

accounting, and physical protection safeguards systems; meet user needs; provide assuranc

of the adequacy of its results; and deal with the requisite spectrum of threats to materi

control and material accounting safeguards. The methodology is planned to be well orient

toward dealing with the types of facilities of interest to NRC. It also is intended to provide reasonable ease of use and good allocation of analysis resources.

*A more detailed description of the material contained in this section can be found in Reference 1

.

164

fa
ar Collusion Methodology

methodology, developed by Science Applications, Inc. (5AI), provides a good analysis

work for integration of material control and physical protection safeguards against

sion (Ref. 4). The SAI methodology consists of analyzing both physical protection and

ial control safeguard components in terms of the individuals who are in a position to

•de their effectiveness. The safeguards components also are represented in terms of

location within the facility. The methodology then analyzes paths from locations

SNM is accessible to the facility boundary, determining the individuals who have

5 to SNM and those who can control the safeguards along these paths. By considering

ndividuals responsible for safeguards components along a path, the conspiratorial group

dividuals who can most effectively degrade safeguards is determined. When all such

date paths for theft are analyzed, the method is designed to locate and list the paths

vulnerable to collusion by any specified group of employees and to list the remaining

jards along the path. Since the methodology considers both physical protection and

ial control safeguards, it provides an integrated safeguards approach to the analysis

cility vulnerability to collusion.

New Approaches to Performing MC&A Operations

asent, there are two programs being sponsored by the NRC and one by the Department of

7 to develop new approaches to performing material control and material accounting

tions. They are studies concerning: safeguards use of process control and quality

-'}1 data, controllable unit accounting, and rapid nuclear material control systems. The

se and status of each program will be discussed briefly.

RC study on possible safeguards use of process and quality control data is intended to ate the potential safeguards utility of such data currently generated by licensees for

jps control and quality control purposes. A short-term preliminary evaluation of this

pt has just been initiated. If this study produces significant results, a more

ahensive evaluation will be undertaken as a follow-on to this initial study.

T RC study on controllable unit accounting is intended to develop a technique for dividing

see production operations into controllable units to improve certain material control

aterial accounting capabilities. The basic concept of controllable unit accounting is

aversion of the measurement uncertainty around a process or a portion of a process

controllable unit) into a numerical measure which represents the degree of account-

.ty control which can be attained with a specific processing and measurement system,

concept makes it possible to partition the process into specific controllable units

j size and throughput depend upon the degree of control required for safeguards purposes.

;Dntrollable unit concept has been computer simulated for a hypothetical high-throughput

:ation plant (Ref. 5). The model indicates that the technique may be quite useful in

;ding rapid checks for material loss. A test application of the concept in an operating

is being initiated.

Department of Energy sponsorship, the Los Alamos Scientific Laboratory (LASL) is Dping a "model" rapid nuclear materials control system, called DYMAC (for DYnamic

ials Control) (Ref. 6). The system, under development at the Plutonium Processing

ity (TA-55) at LASL, consists of a plant-wide surveillance system and a network of on-

.londestructive assay (NDA) instruments interfaced to a computerized materials account-

:y and control system. DYMAC incorporates the following key elements: (a) an in-line

it-line measurement system relying heavily on newly developed NDA instruments to provide

1 assay data at key measurement points, (b) direct automated transfer of data from the

floor into a central computer via interactive display terminals at selected measurestations, and (c) an automated accounting system to provide rapid status on material

fees around "unit processes" which are defined on the basis of process logic and

ial residence time in the process and accessibility for measurement.

jrrent Department of Energy schedule projects a completion date of December 1979 for

cessment of the practicality of the DYMAC system at the LASL TA-55 plutonium facility.

DOE processing plants and laboratories (e.g., Rockwell International at Rocky Flats,

165
Colo.; and the Y-l 2 plant at Oak Ridge, Tenn.) are also pursuing the implementation of

computerized material control and accounting systems similar to DYMAC and on similar

schedules. Also, a few licensees are developing or have implemented rapid, computerizec

material -control or process-control systems in their facilities (e.g., GE at Wilmington.

N.C.; Westinghouse at Columbia, S.C.; and AGNS at Barnwell, S.C.). Therefore, rapid nuc material control and accounting is a useful safeguards concept for the present generatic

high-throughput processing plants, as well as for future plants.

New Techniques for Analyzing Data

There are three current NRC development programs which relate to possible new techniques

analyzing material control and material accounting data. One of these is a program to provide statistical analysis packages to the NRC and licensees for their use in analyzir

material control and material accounting data. The other two programs are a time series

analysis study and a study of the implications of strategic analysis.

The program which provides NRC and licensees with statistical analysis packages is desic

to present, in a series of reports, a set of statistical analysis techniques including a discussion of such topics as the power of the statistical test, the individual confidenc

from a negative test that no loss has occurred, the probable amount and process locatior

the loss for a positive test, and comparisons to other candidate statistical tests. The

tests presented in the reports will not only address plant statistics such as inventory

differences and their associated uncertainties, but will also address the modeling of

specific in-plant processes such as conversion of UF

fi to LLOq. This set of statistical

packages is scheduled for completion by the end of Ft78 ana should prove extremely useft

both licensees and NRC in establishing procedures for analyzing material control and

accounting data, as well as for safeguards analyses of process control data.

The basic concept of time series analysis in the context of material accounting is that

proper data analysis techniques, based upon the properties of a time ordered series of

inventory differences, can provide greater power than the current test of inventory diff ence as compared with its uncertainty in detecting continuing losses of small quantities

material. A number of candidate analysis techniques of this type are under investigate

These include the statistical analysis of cumulative inventory differences (CUSUM), a nu of minimum-variance unbiased estimation techniques Coalman Filtering and Minimum-Varianc

Inventory Estimates), and Box-Jenkins modeling (Ref. 7).

Another complementary approach to data analysis involves strategic analysis. The basic

concept of strategic analysis is to model safeguards in a competitive environment contai

a safeguards defender and an adversary. It formulates the safeguards problem as a serie

actions (moves) by the safeguards defender and the adversary. These actions result in varying costs and benefits to each, depending upon the actions taken. The analysis then( makes use of the theory of games to develop optimum strategies for both the adversary an the safeguards defender (Ref. 8).

Measurement and Measurement Quality Control

In addition to the more sophisticated material control and material accounting methods a data analysis procedures discussed previously, material control and material accounting

detection sensitivity can be improved by simply improving measurement techniques and

measurement control programs. Recent material accounting assessments by NRC and several

studies of licensee material accounting programs (Ref. 9) have indicated that improvemen

in this area are within the state-of-the-art at relatively low costs. NRC has a number

on-going development projects in this area whose FY78 and FY79 products will be the deve

ment of coordinated measurement assurance programs; standardized procedures and calibrat

techniques for nondestructive assay; a safeguards measurement handbook which covers all bulk, destructive assay, and nondestructive assay measurement techniques for safeguards

purposes; NUREG documents on accountability methods and preparation of working calibrati

and test materials; and an empirical study of systematic errors.

i ill

ila %

kt.

166
TASK FORCE RECOMMENDATIONS

b development of material control and accounting programs for special nuclear material is dynamic process of research, testing and rulemaking. Recently a special task force of NRC

acialists completed a comprehensive review of the roles and objectives of material control

i accounting in safeguards (Ref. 1). It recommended goals, development programs and

stem upgrades for strategic special nuclear material (SSNM)* covering the next five years.

the short time I have left, I would like to highlight some of the major recommendations

the Task Force.

e major new thrust recommended by the Task Force is the development of material control

stems to provide continuous cognizance of the status of nuclear material in a facility.

B roles of material control are to provide for the timely detection of material diversion

M authorized uses and locations, rapid assessment of possible losses or alleged thefts,

j assurance that no significant loss of material has occurred during intervals between

/sical inventories. The proposed detection goals are set at two kilograms for plutonium

i uranium-233, and five kilograms for high-enriched uranium. The major goals recommended

r material control include the following:

Provide shift monitoring of processing operations and associated records

Assure close vigilance over material

Ensure that as much material as is practicable is maintained in items or otherwise

measurable and controllable forms

Assign direct responsibility and accountability for SSNM to custodians who shall

maintain effective stewardship over the material

Detect within one shift, with high assurance (>_90%), a loss of five formula kilograms**

of SSNM in the form of items or sealed containers accessible to theft

Detect within a day, with high assurance, a loss of five formula kilograms of bulk SSNM

from any controllable unit within a facility

Detect with high assurance a cumulative loss of five formula kilograms of SSNM from any

controllable unit within the period between physical inventories

Maintain procedures and information sufficient to evaluate, within one hour, process

conditions and material records for possible diversion

\ principal roles ascribed to material accounting by the Task Force are those of after-thet detection, assessment of significant losses of SSNM, and assurance that material is

'•sent in assigned locations and in correct amounts. As with material control, detection

abilities are focused on two kilograms for plutonium and uranium-233, and five kilograms

high-enriched uranium. With respect to inventory differences, action limits of three

•mula kilograms were recommended for reinventory, and five formula kilograms for plant

tdown and cleanout inventory. For the assurance functon, considerable emphasis is placed

the control of cumulative inventory differences to prevent systematic trends away from

o. In addition to the above, the following goals were included in the Task Force

ommendations:

Secure and limit access to accounting records to those who have a need to know

Verify the presence and integrity of all items and containers of SSNM within one working day of receipt of shipments

SNM is pi utoniuns uranium-233 , and uranium which is >20% in the U-235 isotope (ie., highinriched uranium)

.

ormula kilograms = [kilograms contained U-235) + 2.5 (kilograms U-233 + Kilograms

lutonium).

167
Detect within ten days of receipt with high assurance (>_ 90%) a loss of five formula

kilograms of SSNM from any shipment

Monitor cumulative shipper-receiver differences to ensure that these values do not exceed five formula kilograms for any period of twelve months.

Resolve statistically significant shipper-receiver differences during the inventory

period in which they are identified

Detect, with high assurance (>_ 90%) through bimonthly material -balance accounting, a loss of five formula kilograms of SSNM from designated accounting areas

Validate the accuracy of waste discards

Assure the recovery of poorly measured scrap during the inventory period in which it i generated

Establish within NRC a capability for monitoring and analyzing shipper/ receiver and

inventory differences across the licensed fuel cycle to detect and investigate trends

and biases

In the long term, the Task Force recommended that the overall safeguards program be goal

oriented with close integration between the physical security, material control and materia

accounting functions, and that future planning take into consideration the interactions tha may exist in terms of redundancy, diversity and synergism to provide a cost effective syste

of in-depth protection.

NRC staff is currently reviewing the Task Force recommendations and is developing an action

plan for a detailed evaluation of each recommendation. The specific areas for upgrading of

the material control and accounting regulations will be determined on the basis of these

eval uations.

168
REFERENCES

F. L. Crane, et al., "Report of the Material Control and Material Accounting Task

Force," NUREG-0450, U.S. Nuclear Regulatory Commission, Division of Safeguards, April

1978.

M. Maltese, K. Goodwin, J. Schlecter, Diversion Path Analysis Handbook (Prepared for

U.S. ERDA by National Bureau of Standards), October 1976.

Lawrence Livermore Laboratories, Technical Briefing, 1977.

Science Applications, Inc. Technical Briefing, 1977.

P. W. Seabaugh, et al . , Application of Controllable Unit Methodology To a Realistic

Model of a High-Throughput, Mixed Oxide Fabrication Process (Draft report prepared for

U.S. ERDA by Mound Laboratory operated by Monsanto Research Corporation), August 1977.

Energy Research & Development Administration, Division of Safeguards & Security,

"Master Plan," ERDA-76/122, September 1976.

D. H. Pike, G. W. Morrison, G. W. Westley, Applications of Kalman Filtering to Nuclear

Material Control ORNL/NUREG/CSD-1 (Prepared for U.S. NRC by Oak Ridge National Laboratory), October 1 977.

D. H. Pike, G. W. Morrison, "A New Approach to Safeguards Accounting," Oak Ridge

National Laboratory (Published in the Proceedings of the INMM meeting in Washington,

D.C.) June 29-July 1 , 1977.

Kirkland B. Stewart, Minimum Variance Unbiased Estimators of Loss and Inventory

Amounts , Battelle Pacific Northwest Laboratories, September 1977.

W. E. Siri, et al., A Study of Nuclear Material Accounting Vol. I NUREG-0290, (Prepared

for U.S. NRC by Lawrence Berkeley Laboratory), April 1977.

Erick May et al., "Summary Report of Material Control and Accounting Assessment at 4 Major Sites," Internal NRC Draft Report, November 1977 (Confidential).

T. I. McSweeny, et al., "Summary of Findings Evaluating Material Accounting Losses at Four Licensed Facilities," (Prepared for U.S. NRC by Battelle Pacific Northwest Laboratories) November 1977. (Draft)

W. E. Siri, et al., A Study of Nuclear Material Accounting Vol. II and III, NUREG-0290,

(Prepared for U.S. NRC by Lawrence Berkeley Laboratory) April 1977.

169
Evaluation of Analytical Capabilities

For Accurate Prediction of Isotopic Correlation Ratios

by

E. H. OTTEWITTE

EG&G Idaho, Inc., Idaho Falls

ABSTRACT

This study looks at the feasibility of using existing physics

methods and data to verify and/or deduce reactor history as in

safeguards applications. Investigation indicates that adequate

physics methods exist for verification calculations but that they are sophisticated, requiring care to achieve the necessary precision. The accuracy of the physics data (therefore, the

adequacy) will vary with isotope; this imposes restrictions and

necessitates care in the choice of isotopes. Deductive calculations will rely on mostly the same physics methods and data as the verification calculations. Limited study of deductive

(backwards) calculations to date shows that they provide valuable insight into the isotope ratios best suited for the

verification (forward) calculations.

KEYWORDS: Isotope correlation ratios; burnup calculations;

reactor history verif icationj reactor history

deduction; safeguards

INTRODUCTION

The purpose of this study was to evaluate the availability and accuracy of existing

methods for predicting fission product, cladding-mutant, and actinide concentrations and

isotopic ratios in discharged reactor fuel elements, as determined by chemical analyses

after various irradiation and cooling (minimum of three months) times. This type of calculation involving theoretical predictions may be referred to as a forward calculation.

The overall goal is to be able to identify fuel and reactor types and reconstruct irradiation history (backwards calculations) from correlations of key nuclides. Initial attentio:

is centered on the forward calculation: this is the area where significant work seems to have been done. It would also seem to necessarily precede the backward calculation.

Numerous empirical correlations of chemical analysis data exist in the literature.

These are not of interest per se except that they may serve as benchmarks and they do indicate sensitivities of correlations to parameters (like moderator ratio) . Very few

attempts at prediction and subsequent comparison to measurement have been made.

Initially one seeks methods suitable for analyzing individual cases. Later, one migh

develop a data set of predicted correlations for use in recognizing fuel element and reacti

types when unknown. Finally, unfolding techniques could be attempted with such data sets

as bases.

To predict the various mutant concentrations, the methods must

(1) produce suitably - averaged reaction rates and decay data for each

transmutation chain step of interest. This implies a cross-section processor

and a neutronics calculation,

(2) carry out the calculations in some burnup code, preferably with capability for

(a) reading in the transmutation chains of interest,

(b) fuel management (e.g., fuel shuffling, reprocessing/recycling),

(c) adjusting the flux level as fuel is consumed, and

(d) flux-spectrum re-averaging or otherwise getting burnup-dependent cross sections

.

170
The actual mutant concentrations in discharge fuel and blanket elements may be expected

vary with axial length, and location in the reactor. Early cores, for which chemical

ilysis data are already available, have been irradiated on a batch basis. Present core adiations tend to utilize fuel shuffling techniques to maximize burnups and minimize

er peaking factors. Anticipating that most chemical analysis data will be for whole

1 elements, or batches of such, it may be important to include multi-dimensional caplity in the burnup/fuel management (movement and inventory) computer programs; should

al drill-samples or pellets be analyzed, multi-dimensional capability becomes even more

ortant

.

The mutants of specific interest in this study were isotopes 99-104,6 of Ru, 133-5,7

Zs, 140-2,4 of Ce, 142-6,8,50 of Nd, and 150-6 of Eu. Analysis of all major reactor

cepts is of interest, but prime concern must be LWR's whose fuel is already being

rocessed. As BWR analysis tends to require an extra spatial dimension, the specific

nination is further focused on PWR's.

METHOD OF INVESTIGATION

To determine the required and desired features of the theoretical tools, two principal

lods are used:

(1) intuition derived from general neutronic calculational experience

(2) comparison to measurements of predictions from various theoretical methods.

The latter data base is presently quite small.

Several questions come immediately to mind:

(1) What is the uncertainty in the measurements?

(2) What is to be compared? Absolute contents are not generally measurable. Isotopic compositions - e.g., 238Pu/Pu, etc. - are usually measured only for the

element Pu. Certain isotopic ratios seem to be of interest; however, these are generally chosen so as to be linear with burnup and independent of intervening

reactor downtime gaps. Thus, they do not reflect reactor history, only the

cumulative burnup.

(3) What is the relationship of success in predicting ratios which are of interest xe to success in predicting ratios of interest here?

(4) What is the uncertainty in the nuclear data used in theoretical calculations?

These and other pertinent questions are all examined in detail here. Many detailed

;iderations come into play. Advantage is taken wherever possible of related work in the

_:rature.

RESULTS

Sources of Information

Current Activity

Table I lists some recognized centers of related activity. Many of these organizations

together under the umbrella organization ESARDA: European Safeguards Research and

y:lopment Association. Its purpose is to coordinate the research and development work

ited to safeguards carried out by those laboratories operated directly by the European

lission and by other organizations within the European Economic Community.

One ESARDA Working Group has set up an isotopic correlation data bank at the Ispra

tblishment so that information provided by European utilities can be evaluated to pro-

; a better understanding of the correct correlations. The data collection has concen-

:ed on heavy isotope correlations but the working group is encouraging experimental

vities in the fields of fission gas correlations and other fission product correlations,

group is also considering the relationship between input analyses and isotopic correlais using linear regression analysis and statistical tests. This will provide a link

. the Working Group on Integral Experiments. One goal is the selection of a small

er of key correlations containing a maximum of information. Members of the group are assisting the International Atomic Energy Agency (IAEA) in setting up a similar data

171
Table I Centers of related activity

Organization

ENEL

(JRC*)

Location

Ispra

Karlsruhe

Leaders

A. Cricchio

European TU,

GfK, Project

Spaltstoff-f lusskontrolle

(JRC*)

Lothar Koch

CEN, Eurochemie

CNEN

RCN

BNWL

INEL

SGAE

WCAP

ORNL

Mol,

Belgium

Bologna,

Italy

Netherlands

Richland,

Washington

Idaho Falls

Austria

Pittsburgh

Oak Ridge

D. E. Christensen

R. P. Matsen

D. L. Prezbindowski

W. Maeck

W. Emel

K. Goroncek

F. Pichlmayer

G . Falta

M. Lammer

G. Eder

Nodvik

Miller, D. L.

S . Raman

Areas of Interest

Measurements

Theoretical Parameter Studies

Evaluation of FP Nuclear Data Needs

Establish an isotope correlation banl at Ispra.

Study of Correlations in Whole Fuel

Cycle, esp . Transplutonium:

Theoretical & Empirical

Correlations of U & Pu with FP Specti

in MOL IV Use of iso corrlns in MTR fuel repro

BU vs. Eu 154/155 ratio

Corrln of U/Pu w y spectra for BWR's

Empirical isotope correlations, Theo

retical calculations

High precision FP yields and analyti

measurements on fuels, isotope

correlations

"Backward" reconstruction of fuel am

reactor types and history

Saxton Core II, III San Onofre

YankeeRowe V, VI, VII

Actinide XSCS waste production

*The Euratom Joint Research Center is located (jointly) at the ENEL lab in Ispra and

at the Transuranium Institute (TUI) at Karlsruhe.

The ESARDA Working Group on Integral Experiments is concerned with the development

and execution of experiments in which various safeguards techniques are employed on a controlled production campaign over a significant period of time. In general, the main

objectives of such experiments include the evaluation of measurement errors, the testing '• of different instruments and techniques, and the analysis of "operating losses", all

directed towards closing the material balance for an actual campaign.

The Oesterreichische Studiengesellschaf t fuer Atomenergie GmbH (SGAE) , Reaktorzentrui

Seibersdorf has studied backward calculations and their inter-relationship with forward

calculations. In each case the use of a gamma spectrometer was assumed for measuring the

mutant concentrations. Thus, their mutants of interest and their conclusions are not

necessarily ours. In particular they require that their mutants emit y-rays with energie.:

greater than 500 keV.

Reactor Fuels Measured

Existing reactor fuel chemical measurements represent potential benchmarks for verif

cation of calculational tools. To qualify, however, they should include nuclides of

interest and be accompanied by complete model descriptions of the reactor and its history

These are difficult to obtain.

sat

telai

The

laf

£?

SS CO

Itiiin
Table II lists the commercial reprocessors . These complement the organizations of

'.ble I in measuring spent fuel compositions. To date fuels have been mainly reprocessed

the NFS plant (U.S.A.), the Eurochemic plant (Mol) , the WAK plant (Karlsruhe), and the

TL plant (Winds cale)

.

Table II Oxide-fuel-reprocessors

Site

Purex separation

via

Capacity

metric tons U/yr In service

West Valley, N.Y. PC* 300 1966-1972

West Valley, N.Y. PC 600 (total via uncertain

expansion)

Barnwell, S.C. CE/PC 1,500 uncertain

Windscale, U.K. PC /MS 150 1970-73

Thorp, U.K. PC/MS 1,000 uncertain

La Hague, France CE/MS 800 uncertain

La Hague, France CE/MS 1,600 1984-1986

completion

Karlsruhe, W. Germany MS 35 1971-

not yet decided PC/MS 1,400 1985 completion

Mol, Belgium PC 60 1966-1974

Mol, Belgium PC 60 1981-1983

(restartup of

above unit)

Tokai Mura, Japan MS 210 uncertain

Opulsed column; MS=mixed; CE=centrifugal extractor

Theoretical Activity

Much activity in empirical correlation of spent fuel measurements is evident in the

zerature. This forms a solid base of valuable information for many applications, but

es not relate to the goals here except as: (1) a source of possible benchmarks and (2) a ;sible indicator of the sensitivity of an isotope ratio to various reactor parameters,

addition to the empirical activity, some theoretical activity has also been evident at

JL (Washington), GFK-Karlsruhe , and JRC-Ispra. This stems from the desire to extend

rrelations to higher burnups and different fuel types.

The theoretical analysis performed at JRC-Ispra especially deals with heavy isotope

"relations; a wide range of PWR and BWR fuels was analyzed up to a burnup of 40,000 MWd/t,

I the results were compared to experimental data with adequate agreement. For correlal ms concerning PWR fuels, a systematic investigation was made of their dependence on the

.lowing parameters:

fuel enrichment (in the range 2% - 4%) — moderator-to-fuel volume ratio (in the range 1.2-2.2)

cladding material (Zircaloy or stainless steel)

^1 — control system (poison in the moderator or fixed absorbers)

! Similar studies were underway at GFK-Karlsruhe, but with main emphasis on the burnup

endence of effective cross sections which control linear correlations of heavy isotopes . iethod was being developed for the evaluation of cross section ratios from experimental

:a for isotopic ratios which have been observed to be linearly independent.

Existing Burnup Codes and Code Systems

- Table III lists most of the burnup codes which have been encountered. Some include

:1 management. There are far too many to allow detailed analysis of all within the scope

! this study.

173
Table III Some (not all) of the current burnup codes

LOQc IN dlllc Yr Sour c

6

L-OOc IN dill

6

i r Source

ABBRAND IKE HEXBU FINLAND

ACTIN EURATOM HYLAS

AIMFIRE AI IKEV IKE ANALYT IKE ISOCRUNCH

ANDROMEDA 72 JAERI ISOGEN

APOLLO 74 JAERI ISOSEARCH

ASB 71 73 KFK ISOTEX-1 74 ISPRA

ASSAULT ORNL ISOTOPES

BABS UK JOSHUA 68 UK

BASS UK LASER 65 West

BOXER 74 EIR LASER/HIC

BRUTUS BOLOGNA LEOPARD West

BUG2 LEOPARD/ SPOTS West

BUPO 71 UK MAFIA1

BURP 2 66 MANDI KFK

BURSQUID 72 ENEL (IT) MEDIUM 71 SIEMENS

CACA-2 76 ORNL NEUTRA 72 GEESTHACHT

CAPHE 75 FRANCE NUCY ORNL

CHAINS ORIGEN 73 ORNL

CINDER LASL PABUC AI CITATION 71 ORNL PANDA 71 WEST.

CLUB 72 ORNL PDQ7

C0ND0R3 70 ITALY PDQ8

CRACKLE 7 3 UK PENELOPE /CITAT. 75 SPAIN

CREAM7

71 UK PHENIX 69 LASL

2DB BNWL RAD2

3DB BNWL RAPFU

DBUFT11 REBUS

2

74 ANL

DDB REST 72 KFK

EDDT RIBD UNI

DELIGHT2

74 JAPAN RIBD-II 75 HEDL

DTFBURN RLT-4

DWARF 72 INEL RSYST IKE 3DXT/DEP3 SINBAD 69 INEL

FABGEN 75 ORNL SIZZLE AI FARED SPECTAR 74 YUGOSLV.

FEVER7 SYN GE FISSPROD-2 75 AECL SYNBURN 76 ANL

FLARE 64 GEAP SYNTRON 72 RISOE

FREVAP6 SYNTRONVOID 74 RISOE

FUMBLE TRIBU 73 BELGIUM

FURNACEJ 72 JAERI TRISIC 73 ARGENTINA

GAD, REV. TURTLE 24.0 71 WEST.

GAFFE VARMU-1 BELG.

GARGOYLE EIR VENTURE-BU ORNL

GAUGE GA,EIR XPOSE (LEOPARD) 75 EXXON

XTG 75 EXXON

Existing Comparisons of Theoretical Calculations with Measurement

Few comparisons have been reported in the literature. Of those which do exist, the calculations are generally either zero-dimensional or 3-dimensional . There is little ir between. The zero-dimensional approach appears to suffice in asymptotic flux regions (e from core boundaries or control rods), but is inadequate elsewhere. The limited 3D calc lations made so far have been compared to measured axial distributions with reasonable

success. Somewhat strangely, integral comparisons from whole fuel elements or groups of such have not been found yet in the literature. Comparisons to date are listed in Table

174
List of existing

Table IV

comparisons of calculation to measurement

jthors

. Ariemma

. Bramati

. Galliani

. PaclettiGualandi

. Zaffiro

. Cricchio

. Koch

N. Fox

Lab

ENEL (Ispra)

Euratorn

(TUI)

Winfrith

Reactor Calculational

Reference Fuel Methods

Reaktortagung Garigliano 2D neutronic & 1971, Bonn (BWR) BY EQUIPOISE,

EUR-4638 SQUID

(1971) 3D FLARE feedback

RIBOT cross sections: 2G,5G,2K,

10K mesh

1971 BNES SGHWR a: WIMS (69G->7G)

,

Conf at METHUSELAH

Canterbury <j>: DSN(7G), ID cell calcs.

Feedback: JOSHUA

II

Measurements

Ft* by 137Cs Y~ spectrometry

8 3Kr/86Kr and 84Kr/

83Kr by mass spec

anal of pellets

Actinides and

11+8Nd by mass spec

Actinides

Davis, Jr ORNL

W. Kee

C.A. Vaughn

L. Tobias

ORNL/TM-5691

(1976)

Peach

Bottom

HTGR

FTE-4 Fuel

Rods(2)

ORIGEN U isotopic composition from head-end

reprocessing y -s can- ning for 85Kr, 90 Sr,

95

110 144

Zr, 95Nb, 106 Ru,

Ag, 125 Sb, 131+ » 7 Cs,

Ce,

152,i+,

5

Eu

England

ggi Ispra

Calder ARGOSY vs Hall WIMS-D vs WIMS-D

IAEA-SM- Garigliano point burnup

201/44 Trino Ver- ISOTEX-1, GGC-2

cellese VAK

240.1 Pu

FT , 81+ Kr/ 86Kr, 86 Kr/

lt Kr/ 83Kr, 132 Xe/ al

131+ Xe, 131+ Xe/ 131 Xe,

132Xe, 131 Xe, ll+6Nd/

H+ 8Ndj 148Nd/ 148Nd)

145Nd, 11+6 Nd/ ll+5Nd,

15137

'Eu/ 137Cs 134Cs/

Cs

DCODE (ID-TT)

RIBD(FP)

238-42 . McN eece United Nuclear UNI— 436 N Reactor! Pu

Industries, (1975) Mark IV, Inc. Richland Mark 1A

Wash. Fuels

[ is atom % BU

On the basis of existing experience with prediction of actinides and a few burnup indi-

:ors, multi-dimensional techniques appear to be needed for computational accuracies better

an 5-10%. Whether this is needed to predict fission products mentioned in Section 1.0,

secially their ratios, remains to be seen. It may not be worthwhile to use a < 10%-error

:hnique with nuclear data containing > 10% error.

Problem Understanding: What Are the Mutant Parameters of Interest?

Given the goals described in the introduction, the question arises - just which mutant

ameters are of interest? More specifically, is there a complete or near-complete basis

: of parameters which will sufficiently characterize a reactor and its history. This

ses more questions:

(1) What, specifically, do we desire to know about a given reactor fuel?

(2) how does one go about finding the basis set7

175
(3) can one select just a few parameters (the ones with the most Inherent accuracy'

and find permutations through a Gram-Schmidt type of orthogonalization process

which will then give a "complete" set? Hints at the feasibility of using such

permutations are already evidenced in existing papers. Such a complete set ma]

form a basis set for the "backward" unfolding techniques also.

What Do You Really Want to Know About a Given Reactor Fuel?

A basic question arises at the outset - are we interested in general reactor behavi< or are we looking for some local substitution or diversion? By general reactor behavior mean abnormal conditions like downtime, reduced rating, overpower, and excursions. By

local behavior we mean:

(1) Was any irradiated fuel element replaced other than as scheduled?

(2) Has any fuel been diverted?

(3) Were any special samples introduced?

This and other similar questions should show up as deviate chemical results in a batch o: elements. One might however have to specify the limit to the number of elements in a chemical dissolution batch.

The significance of the local vs general question is in the complexity of the nucli

model. The answer to the question is that we are interested in both. However, it is wo: while to note that the local solution requires considerable more complexity in the model

than that for the general reactor. This is discussed further in the next section.

Determining the Complexity of the Nuclear Model

In the case of fast reactors, axial and radial blankets will also be present. TabL

shows the calculated equilibrium fuel discharge composition for a 1000 MWe LMFBR. As cai be seen, the isotopic weight fractions vary considerably among the different core and

blanket regions. Note particularly how different are the "Isotopic Weight Fractions in

Each Chemical Element" for:

(a) (inner) core Zone 1 vs (outer) core Zone 2.

(b) inner radial blanket vs outer radial blanket

(c) individual zones vs "total reactor".

These results were obtained with a one-dimensional neutronic calculation plus burnup and

fuel management routines. Runs were separately made in both the radial and axial direct

As the axial blankets are a part of the fuel element, they may be expected to be mi: in with the core in the chemical analysis. The radial blanket, however, is physically

separate. By excluding the explicit calculation of radial blanket concentrations (just j use a transverse buckling), a one-dimensional axial calculation may suffice. Indeed, it

doubtful that the radial blanket concentrations can add significant information to that

the core in characterizing the nature or history of the reactor. One question remains j however: Do we need to check the radial blanket to see if special samples have been int,

duced or if diversion has occurred? In those cases, we are probably back to multi-dimen

al calculations.

For LWR's one can divide the fuel elements into two categories:

(1) those in an asymptotic flux spectrum,

(2) those in a perturbed flux spectrum.

This is quite important. By asymptotic we mean as in the cell of a large lattice, far

from control rods or enrichment zone boundaries. The latter corresponds to perturbed fl

situations. In practice, perturbations are always present due to reactor finiteness, bu

they are frequently insignificant. This is particularly so for earlier cores which are operated on a once-through batch basis. Current cores tends to be divided into radial f shuffling (management) zones of differing enrichment, thereby introducing more boundarie

within the core region.

The importance of this all is that for asymptotic flux spectra, zero-dimensional

techniques calculate isotopic concentrations well whereas perturbed flux regions require

one- or more-dimensional representation.

176
Table V

Equilibrium fuel discharge composition, 1000-Mwe LMFBR advanced design

For safeguards considerations, however, there is no doubt that one must calculate all

"R core regions and all LMFBR core and blanket regions: not for reactor type and history

entification purposes, rather to check that no Pu has been diverted. Thus, we see that

e purpose may need only a simple calculational tool; another purpose requires an elabate tool: all for the same reactor and history.

Some of the early reactor cores have been reprocessed and analyzed. These fuels were

radiated on a batch basis with no fuel shifting. Even where fuel-shifting management is

heduled, the first batch of elements out of a reactor will have been exposed only at one

actor location. However, as it is very attractive to shift fuel to increase the

rnup while simultaneously minimizing peaking, one must assume this philosophy

predominate for the future. This can be readily accommodated in a forward calculation,

backwards calculation would seem to be very hard to accomplish if any fuel shifting was

neduled. The direct approach would be to incorporate fuel shuffling into a backwards

story model. This would add one or more additional parameters, however.

Another potential problem area is recycled fuel (more of a problem for LMFBRs than

r LWRs) . There the feed compositions may change with time. These can depend on such

Lngs as reprocessing techniques: some labs may reprocess blanket Pu separate from core

, while others mix the two. This will significantly affect the Pu isotopic compositions

the feed.

Different reprocessing techniques also affect the spent fuel analyses. WAK (Germany)

is a chop and leach process which homogenizes fuel. This has the effect that the isotopic

;tribution of fission gases is constant throughout a dissolution.

How Might One Go About Deducing Reactor Type and History?

We assume we are confronted with one or more black box (no apriori knowledge) reactor

.cents. A remote gamma scan can give an axial radiation profile to tell us if there are

' heterogenities : graded enrichment steps or blanket zones, or special materials. Also,

will indicate fuel column length, relative power distribution, and fission product

ration. Neutron radiography can tell of local hydrides, H2~containing impurities, fuel

tensions, and fuel structure.

Assuming homogeneity, the element is next presumably dissolved and the inquirer is

sequently presented with the chemical analysis. Mass spectrometry subsequently reveals

isotopic analysis. Repeating both the chemical separation and the mass spec analyses

177
at later times would give time-dependence as well. Alternately, this may be feasible thr I gamma spectrometry of small "thin" samples.

We now would attempt to characterize the reactor. Actinide composition should speci

the original kind of fissile and fertile fuels present, and the approximate neutron spect

Combined with the differences in fission product yields per fissile isotope and spectrum

one would try to deduce the original composition. Possibly the condition of the cladding

outer surface and its composition can help to identify the reactor coolant. Coupling the

above with the residual compositions of cladding, fuel, and blanket (if present), one sho

be able to pinpoint the type of reactor including the gross characterization of the neutri

spectrum. Isotopic composition and fuel diameter can distinguish between a PWR and a BWRj

With that, an accurate spectrum-averaged set of fission product yields is obtained for ea fissionable nucleus of interest. This is very important to determining the original comp j osition and time-dependent history.

Many isotopic ratios are strongly dependent upon enrichment, fuel-to-moderator ratio

j

and fuel pin diameter. Weaker dependences on clad and control method may also be utilizaj

This should help determine the reactor composition and associated nuclear data base. Fir! through ratios of isotopes with a variable array of half lives one can try to reconstruct

irradiation history, especially in terms of deviations from full-power operation.

Mutant Parameters for Backward Calculations

Note that the following has been derived from a program which relied on a gamma

spectrometer for measuring mutant concentrations.

The selection of mutants to be used for interpretation is guided by the following

considerations

:

(1) Accuracy of the nuclear data;

(2) Accuracy of the measurement;

(3) A range of half-lives among the mutants so as to yield information about

different periods of the reactor history: irradiation times and shutdown

times. Most fission products will only yield information about the later

part of the irradiation history.

(4) At least some mutants with significant cross sections are desirable as that

makes them sensitive to down times.

(5) Fission products must have adequate yields from fission.

(6) Burnup may be inferred from long-lived fission products. In a gamma spectromet

program the following conclusions were made.

137Cs (30 yr.) is far superior with respect to the amount of additional

information needed for interpretation: it requires no knowledge of neutron

spectrum or fuel compositions and almost no knowledge of irradiation history.

However, one must be sure that no significant loss (<10%) of 137Cs has occurrec- Also migration of 137Cs in very bulky fuel elements could cause errors if smal]

local samples were taken.

1 06 Ru - 106 Rh (1.0 yr.) is next best with respect to half-life, but has

much higher fission yield for Pu fissions than for U fissions.

11+lt Ce -11+lt Pr (285.4 d) is only useful if the irradiation history is eil known or rather short.

95Zr - 95Nb (65.5/35.15 d) is only useful for very short irradiations.

131+ Cs is also used with 137Cs as a ratio. , (7) Cooling time can be determined from measured activity ratios of fission producl

of various half-lives if it can be assumed that they were in saturation

immediately before shutdown. Figure 1 shows some typical curves.

(8) Methods to determine flux t$> and irradiation time T require a constant neutron i or a rather accurate knowledge of the neutron spectrum. The 13i+ Cs/ 137Cs activ:

ratio gives the integrated neutron flux with low accuracy.

(9) Methods exist to determine the fraction of Pu fissions. The fission yield for

106 Ru - 106 Rh varies by an order of magnitude between U and Pu fissions. lttl+ Ci li+ifp r ^ s a fiss i on product of comparable half-life and its fission yield is ra independent of the type fission. By using the lhk Ce -

11+lt Pr/ 1 06 Ru - 106Rh rat: a first order cancellation of decay correction errors is achieved. The activi

ratio decreases very slowly during cooling with an effective half-life of 3.3

178
Figure 1, Dependence of activity ratios of

several fission products on cooling time.

years; therefore, only an approximate decay correction is needed.

Similar attempts with 11+0 Ba -

ltt0 La/ niAg or lk0 Ba - 140La/ 156Eu ratios

failed.

The mutants may be nominally divided into three classes:

a. Primary fission products (P) : all precursors have relatively short

half-lives

,

b. Secondary (decay) fission products (S) : immediate precursor was a Primary.

c. Secondary (absorption) fission product (A): obtained by P(n,y)A reaction.

Now in principle each mutant FP is a function: f (BU,t,A,a,<)>) where:

BU burnup

t time

> X decay rate

o destruction cross section fairly well known

-' flux

This assumes there is only one process leading to formation of the nuclide. Although

tivity to each of the parameters (BU,...<j)) is needed through one nuclide or another,

also desirable to limit the sensitivity for each individual nuclide. For the case

I, if a = A = 0, then P = P(BU) only. For S, one desires ap = 0 = X s = a s such that

(P, Apt) = S(BU,Apt) only. For A, one desires Ap = 0 = \& « o"a, such that A = p$t) = A(BU,ap

<{!t) • Now, in principle, we have P(BU), S(BU,A

p

t), and A(BU,a

p

<pt) , which

be orthognalized to fi(BU) , f2(t), and £3(1^) .

One parameter (BU) deductions are straightforward and frequently done. BU is in

ce,

T /

a

f

<j>(t) dt

sberger and Bruneder* showed how to deduce several parameters. First, they devised

-parameter model: .

'J. Higatsberger and H. Bruneder, Acta Physica Austriaca 28, 94 (1968)

179
With it, forward calculations yielded typically Figures 2 and 3 for 103Ru and 137Cs.' sider the case where [ 103Ru] = lxlO 15 and [ 137Cs] = 3.16xl0 16 ([] denotes concentrati

at time T. Upon overlaying the two curves, these two lines of concentration intersec

T <v 2 years and <j> * 4x10 19 n cm

-2

f > < p f 7 a j

years

-1

4 5S7i»10 1112 13i4iSt

Irradiation Time (years)

Figure 2. - 03 P.u buildup (short-lived primary

fission product, 40 days}.

rradiation Time (years)

2 13 H IS 16

Figure 3. i37 Cs buildup (long-lived primary

fission product, 30 years).

Instead of overlaying curves one could parameterize [ 103Ru] as f(t,<f>), a three-dimensional space. Then by some computerized mathematical technique th

of the * S3Ru and the 137Cs surfaces would (hopefully) uniquely and sharply de

values of T and (f>. To maximize the sharpness (the angle between the surfaces

intersection) , inspection of Figures 4-5 shows that th should vary as much as between different P. Also an "A" surface is quite different (Figure 5).

surface III e inters :1 termine [pi at thej

possibjjj

I

III

Irradiation Time (years)

Figure 4. 13 *Cs buildup (n,i product of l33 Cs,

2.3 years).

Irradiation Time (years) Figure 5. S5Nb buildup (short-lived secondary

product of 95 Zr, 35 days).

180

eit

It

k
j In similar fashion, one would hope that the intersection of the f(X-^, X2 , X3) (volumes

rour-dimensional space) for three isotopes would sharply and uniquely determine the

uneters X]_, X2, and X3. However, we note complexity was rapidly increased just in going

1 one parameter (BU) to two parameters (<f>,T).

Actually, Higatsberger and Bruneder present a formalism for four and more parameters,

iits practicality remains to be demonstrated. In each case one has to assume a historical

1 cr pattern. For 4 parameters one possible model has a continuous "load availability"

ern:

parameters are then T, <f> , n, I. Desirable isotopes for analysis would presumably

ude some with time constants (A or a<j>) of the order of tj, t^, and T.

The problem with the above is that Figures 2-5 are for a point reactor (reactor average).

2 {df} for the fuels and {ay} for the fission products are involved, the analyses are

con spectrum dependent; and the spectrum is spatially dependent.

One solution to the dilemma would be spectrum-independent ratios of isotopes. This

.^De feasible for thermal reactors where Oy for most materials behaves as Oy(E) = However, the heavier fission products tend to be somewhat deformed resulting in

. sr level densities; that leads to some thermal region resonances (non-l/V behavior).

£p Another path is to find additional isotope ratios which indicate that the spectrum is from

:- turbed region (near a boundary or control rod); a suitably-determined correction to the

ijttrum-independent" ratio would then be made . As additional urcertainty would be

Dduced, the "E-Jj" solution (above) is to be preferred.

What Mutant Parameters are Available

This study has assumed that chemical analysis data are the most accurate and reliable,

.s therefore predicated on this. Accuracy will vary with chemical element and depend

the composition: e.g. , two chemical elements may be hard to separate from one another,

•mplete recovery of a given chemical element is infeasible, results become relative,

'pic composition in a chemical element can be determined through mass or y spectrometry.

Thus, the most accurate information will be for ratios of two isotopes whose atomic

stries evidence similar high recovery rates.

1 In summary, we see that we must concentrate our methods on the information from

pe/isotope ratios (^X/^Y) , and isotopic compositions ( AX/X) . The latter should be the accurate.

It is also important to know the relative yields of mutants in typical spent fuel

ses, i.e. , just what is available. Table VI lists the few isotopes which have been

fically measured in reactors for isotope correlation purposes and such. In addition

e above, ll* 8Nd is frequently measured for burnup determination. It is popular because,

1 the other fission products, it is the only one presenting the following main

cteristics

Its cumulative fission yield is practically identical for uranium and plutonium and

is independent of neutron energy.

181
Neodymium and its chain members do not change location in the fuel by diffusi

other transport mechanisms during irradiation.

It is stable and neutron capture by the chain members is practically negligib

compared to their beta decay. Thus it is rather insensitive to neutron flux

changes and shutdown periods in the fuel irradiation.

Its chemical separation and analysis by isotope-dilution mass spectrometry is

relatively simple.

Table VI

Isotopic measurements reported in the literature for spent fuels

Reactor Measurements

Latina (gas-graphite)

Chapelcross (gas-graphite)

Garigliano (corner and std. rods, BWR)

Kahl (BWR)

Gundremmingen (BWR)

Trino Vercellese (PWR)

Dresden (BWR)

Humboldt Bay (BWR)

U, D 5

, Pu

25, 28, 25

1 32 131

83131

Xe,

Xe,

131 132

Xe,

Xe,

241pu

241 Pu,

8i*Kr,

Kr, D 5 , Ft Xe, 132 Xe,

Pu, U, 131t Cs,

Pu, U, D 5 Pu, U, D 5

241

137

Pu

Cs, 131 Xe, 134 Xe

F t = fractional heavy atoms burnt

D5 = fractional 235U depletion

Sensitivity of Mutants to Reactor Parameters and Conditions

The fractional composition of a given Pu isotope can fluctuate 10-20% about a no value due to parameters such as:

(1) moderator-to-fuel volume ratio (MFR) which ranges from 1.2 to 2.2

(2) 235U enrichment (ENR) which ranges from 2-4%

(3) fuel pin diameter (OD)

Reactivity control, e.g., boron concentration, and cladding (zircaloy vs. stainless s affect composition only by a few percent.

For the specific isotopes Np-237 and Pu-238 the atom densities are dependent as [Np-237] = f (1/MFR, ENR, 1/OD, clad (weakly))

[Pu-238] = f (1/MFR, 1/ENR, OD, clad (weakly))

Figure 6 from Foggi and Frandoli (EUR 5071) graphically illustrates some sensitivitie if

Tables VII and VIII list some key parameters for existing PWR's and BWR's. Note;! consistencies. Thus, within a given class (PWR or BWR) the most sensitive parameters It not vary considerably. This should help to distinguish PWR from BWR fuel.

As one might expect, any recycling of Pu leads to an increase in the production II actinides with A>242, which should not be very different between the two LWR types. I'm an enormous accumulation of Cm-242 in later TRINO VERCELLESE samples was caused by a if shutdown of the reactor, which consequently led to an Am-241 buildup of Pu-241 decay.

if Other similar tell-tale decays should be present both in the actinides and in fissioril

products as well.

Nuclide concentrations are inherently sensitive to cross sections, flux, and tin).] The cross sections and flux are, in turn, sensitive to numerous reactor parameters,

of this sensitivity can be removed by considering ratios of one isotope to another; jj numerous ratios are found to exhibit linear dependence upon burnup. This can be useif

that it reduces dependence to one or two variables which can be accurately detenninec

the other hand, these same linear ratios may be now independent of flux level and tir! (irradiation history information). Thus, a combination of concentrations (or isotop: )J compositions) and isotopic ratio information should be of interest here.

182
Pii/U (weiqht

D.

ratio)

MR = 1.2

MR = 1.64

5 HP = 2.2

poison in the n1 leakage of neut

sderator

rons

enrichment

tw/o)

Figure 6. Pu/U to :>5U depletion ratio for different

lattices and different reactivity control methods

(Zircaloy cladding).

Simvi l.VhLl A BM-.-

Kx-I Hi crier Moderator/ \rter Hegi'-n keaitor tuel fc*ti" Equilibria

Beaver allev 0.36*9 1.67 3.2 3.2 Calvert Cliff* (1.382 3.03 Connecticut

Yankee 0.383S 1.515 3.67 frystal River-3 0.34J 1.68 2.90 2.?4 Donald Took

Diablo Canyon1

0.36O9 1.67 2.67 Diablo Canyon -2 0.3669 1.67 3.70

cvn Calhoun O.MS 1.60 2.9S 3. OS

Indian Point-2 0.3669 1.67 2.61 2.92

Indian Point3

0.3669 1.67 3.20

(ewaiaiee 0.36o9 1.60 3. IS 3.38 Main Yankee 1.61 3.01 Midland1

Q;3K 1.6S I.M

O.UtZ 1.79 2.77

2.47 Oconee-2 0.362 I.6S Oconee- 3 2.94 Oyster Crock-: 0.362 1.68 2.90 Palisades D.3S9 1.70 2.74 2.74 Print Seach-l W/ 1.60 2. SO Point Beach-2 0.3669 1.60 2.80 3. OS Prar»e Island1

0 3669 1.60 3. IS 3.38 PTane Island-: 0.3669 1.60 3. IS 3.18 Robinson2

0.3*69 1.675 2.73 RussellvUle 0.362 1.6* 2.90 2.94 Sale*-! 0.3669 1.6" 3.20

Seqisia1

0.36S9 1.61 3.20

3.38 Surry1

0. 3669 1.675 2.94 Surry -2 C.3669 1.675 2.94 3. SB Three * le :.94 Island 0.362 1.67 2.90 Turkev Point3

0.3669 1.67 2.73

Turkey Point-4 0..V9 1.67 2.73

Zion-1 0.366? 1.72 3.20

lion2

0.366? l„!»l s.;o Tanke* ftowe i.:«*> J. 41 Core I 4 II

factor

Arnold

Bell Bif Rock Fvint

Brown' s Ferry1

Brown's Ferry2

Brown 1 s Ferry - 3 Bnnsvick 1

Dresden1

Dresden2

Dresden3

Fitmatrick

Hitch LaOosse

Millstone Point

Mcnti cello

Nine Kile Point Oyster Creek1

PlLrrtji Quad aties I I II ShoTenan Vermt Tankee

Peach Bottoar-1 Peach Bottm-2

Peach Bortoa-3

0.483

i}. 488 0.487

0.3S0

0.448

0.488 0.488

0.484

0.488

0.488 0.488

0.488

0.488

0.488

0.488

2.19 2.2S

2.15

2.36

2.36 2.23

3.63

2.07 2.03

2.10 2.10

2.18 2.18

Z-28 2.29 2.15 2.19

2.19

Isotopic Ratios vs> Isotopic Compositions

- The buildup and depletion of a given isotope is dependent upon exponential terms

" and e~^ t

A = decay constants

t = time

a = microscopic spectrum-averaged transmutation cross section

<j> = neutron flux

183
In as much as all theoretical calculations are to be compared to measurement, it is imp

to recognize that considerable uncertainty lies in the physical quantity <f>. As illusti

consider a reactor fuel reportedly exposed to a burnup of 30 MWd/kg U. This is probab]

uncertain by at least 5% just due to uncertainties in knowledge of exact calorimetric p levels of operation. Now analytically burnup is just the integral

/

a,. <J> N£ . dr dt. riss fuel —

Thus, uncertainty in the power levels results in similar uncertainties in the deduced i (°fiss an(3 Nfue i are rather certain) . Transmutation is generally proportional to term

like:

[l - exp (-a((>t)] (one-step process)

If a ^ 10 barns = 10 23 cm2

} . .

. „ -,^11+ _? • _i I typical power reactor

<j> ^ 10 n cm 11 sec 1 \ , . . .

1 year)

j

. „ in7 ,~ conditions

t ^ 10' sec (1/3

such that a(J> is in the neighborhood of 0.01, then

[l - exp (-a<}>t)] ^ o(j)t and production is linear with burnup (outfit) As example, for 238U-> 239Pu

4928N

< °28

o For a two-step process 238U •* 239Pu *

2Lf0Pu

/ n, Y \ / n, Y \ ti s2

2iN0 * \

CT 28 > \°49 )

If one now takes ratios of mutant production, <j>t frequently cancels out. Thus, ratios

much more dependent on the a" (t) involved than on the precise history. Thus the ratio

(Pu/U) /

(

2 39Pu • 2l+2 Pu/( 2l+0 Pu) 2

) , employed in the correlation literature, reduces to constant times (<(>t) / (<$>t • [^t]^ /[4>t] 2

) 2 = 1- For fission products as previously classif

"P" = <f>t, while "S" and "A" are of the form:

S(t) = a(t- [l - e~ p

1"] ) . a = constant

Xp

For small t, S(t) ^a 4p t

2

. For A, Xp becomes a <J>.

The {o'} are strongly dependent upon spectrum which is affected by moderating rati

enrichment, and pin diameter, and to a lesser extent on cladding and leakage control

mechanism.

In summary, isotopic compositions will depend on both flux and cross section. Fl should generally be the more uncertain, especially, in trying to relate to experiment

values. Ratios eliminate problems of flux dependency, but carry little history infom

Preliminary Choice of Mutant Parameters

Parameters involving two groups of mutants quickly come to mind: the Group I act for which much experience and data already exist, and the fission products of original

interest in this study because of the high accuracy with which they can be measured.

Before an accurate and general approach to the problem of constructing or choosii

isotopic ratios can be defined the important question of the sensitivity of an isotop:

ratio to diversion (or other unreported behavior) must be examined. Any one isotopic

will have zero sensitivity to certain types of diversion; this zero sensitivity is ca.' a "blind spot". Methods for identifying and compensating for "blind spots" in the

isotopic ratio require the solution of a system of equations which predict the possih

combinations of overall isotope changes that will produce no response in the isotopic

ratios being used. The use of a number of distinct ratios that is greater than the n'

:

It!

lis

'$

Mi

fel

184
r isotopic concentrations that the diverter may conceivably manipulate insures that the

ilind spots" to diversion are covered. Two or three ratios may be all that are required

practice since other constraints may eliminate many mathematically-possible but

ctically- impossible diversions.

Problem Understanding: Considerations in Comparing Theory to Measurement

Several factors may adversely affect the comparison of theoretical prediction to

! eriment: First of all, there exist uncertainties in both the theoretical calculation

, in the measurement. Second, there is the question of whether to even consider some

-central reactor zones. Third, problems of mutant migration (diffusion) can occur.

:ally, there is the question of how to relate success in predicting ratio A/B (done in

literature) to success in predicting C/D of interest here.

Measurement Uncertainties

Measurement errors may be expected to arise in chemical separation and analysis, and

the mass spectrometric analysis. In part the uncertainty will be due to different

nical recovery efficiencies. In the Diablo Canyon analysis appreciable Xe, Kr, I, U,

Pu were lost.

Isotope migration effects . For some fission products axial and radial migration can

ar. One underlying mechanism can be simple diffusion from a hot region towards a colder

In the well known case of Cs , this is probably due to the gaseous and chemically

rt nature of its precursor Xe and the tendency to diffuse from a hotter temperature

irds a cooler one. Other elements with low chemical affinity may also diffuse. (The

i other case with an inert gas precursor is Kr -> Rb.)

To avoid this problem in general it will be better to always analyze the whole fuel

aent. In the specific case of Cs, it will also be important that the cladding integrity

not violated resulting in loss of Xe gas through it.

Reported measurement uncertainties: Table IX. In the analysis of Saxton Core II,

:inghouse used mass spectrometric analysis to determine the U and Pu isotopic distribui and the 239Pu/ 238U and 11+8Nd/ 238U atom ratios in each spent fuel sample. J37Cs was

;ured radiochemically . Alpha spectrometric and radiochemical analyses were made for

•u, 238Pu, 237Np, 21+1 Am, 2tt3Am, 242 Cm, and 21+lt Cm.

HEDL and ARHCO analyzed the same samples and reported their uncertainties. A German

prt gives the measured uncertainty in uranium enrichment. Karlsruhe and Ispra both

yzed the same sample for two fission products. The difference in their results is

:-.n as a measure of uncertainty.

Table IX summarizes these data and makes some rough estimates of the state of the art. conclusion is that the standard "plutonides" are quite accurately measurable. Exotic

nides and fission products are less so. The table is somewhat misleading for the

ion products as a ratio is not involved. Presumably, with proper choice of elements,

lical uncertainties can be virtually eliminated leaving only those from mass spectrometric

ysis

.

Theoretical uncertainties: Computer precision

;j The length of a computer word is 32 bits for IBM, 60 bits for CDC. Each word is did into a mantissa and an exponent, which give the precision (e) and range (R) of the

ers representable. For IBM this is:

e T = 10- 7 10" 78 < Rj < 10+75

' CDC,

e = 10-15 10-294 < R < 10+332

i

i c c

185
Table IX

Measurement uncertainties

(c)

German

(%)

W(a)

Pavamo r cLL CXillC for Lei HEDL

Pu-238/Pu 2.3 8.8, 3.4 10, 9.8

Pu-239/Pu 0.03 0.1, 0.06 0.3, 0.06

Pu-240/Pu 0.2 0.53, 0.5 2.1, 0.13

Pu-241/Pu 0.3 0.42, 0.5 3.4, 2.8

Pu-242/Pu 0.9 3., 3.0 7.4, 3.3

Pu-239/U-238 0.7

Np-237/U-238 15

Cm-242/Pu-239 10

Cm-244/Pu-239 20

U-234/U 29.4

U-235/U 0.9

U-236/U 5.6

U-238/U 0.01

Cs-137 8 Nd-148

Karlsruhe

vs. Ispra

(d) Estimated 2 Range (%) 2-10

0.03-0.10

0.2-0.5

0.3-0.5

0.9-3

0.7

1.08

15.

10.

20. 29.

1.

6..01

(2)

(a) WCAP-3385-56 (pt.2), uncertainty is 2a.

(b) UNI-436, uncertainty basis not given.

(c) uncertainty basis not known.

(d) la estimate from comparison of two different measurements on the same sample.

In examining computer programs we must consider various sources of error. For a burnup code, ordinary differential equations are generally solved by finite difference

techniques. This implies several sources of error:

(1) (truncation) error in the finite difference approximation to a differential

operator. This depends on the number of terms in the formula and on whether

is implicit or explicit. Generally, implicit formulas have less truncation

error than corresponding explicit formulas; however, implicit formulas usuall

require iteration, and are thus prone to propagation of roundoff error, roundoff error arises from the finite nature of the computing machine which c only deal with finitely represented numbers. This is expressed explicitly in the word sizes (in bits) , and the number of those bits devoted to precision ( range) . The greatest loss of significance in numbers occurs when two numbers

about the same size are subtracted so that most of the leading digits cancel

Unless care is taken in advance, this can happen almost any place in a long

computation.

instability (error propagation). Frequently, numbers produced at one stage £ fed back into the computer algorithm to be processed again and again. The

question then arises - will a small error grow or decay through the successi\

iterations.

range errors - exceeding the limits of the number system. These are not gene encountered. This arises from successive multiplication of small or large m

so as to exceed the range provided by the leading bits in the word.

In considering nuclde buildup and depletion, one often encounters terms like:

(3)

(4)

N2(t) = Ni(o)a2

[e _6 2 t "It]

where a, 3 are production and destruction coefficients of the form a<f> or A,

For t small enough this simplifies to N2 (t) = (Xot Mo) J2 '

For a multiple step process, e.g., 2-> 3-> 4-*- 5

: (a 2 t)(a 3 t)(a lt t)(a 5 t)

186

( n a.)t

i=2

1 m-1
us burnup calculations involve: (1) the difference between small numbers, and (2) the

oduct of numerous coefficients in a long chain process. In addition care must be taken

th units: for (m-l)=6 one must calculate U(a <j>) in units of sec

- -'- and not

3~24)6 x n (cj-^Cb) cj>) . They both give the same result in theory, but the second, which

lays the correction for units, would give an incorrect result on an IBM computer for

-1) _> 6. Similarly, II (a^) x <j>6 could also easily exceed the IBM single precision

age.

A realistic example is some calculations made on the IBM 360 computer in single

acision for the transmutation scheme fihown in Figure 7. Most concentrations such as

-243 build up smoothly (Figure 8) due to high concentration, direct paths, high fluxes,

i appreciable cross sections: core regions "a"-"E" (see figure). In the remote (low

jx) blanket regions abnormal behavior is noted for distant chain members like Am-243.

Atomic Weight

232 233 234 235 236 237 238 239 240 241 242 243 244

Atomic

Number Symbol

92 U

XI Reaction —-

n.2n Reaction —

S" Decay J a Decay ^

S Starting Constituent • Accounted Product

Figure 7. Branches of Interest 1n F8R fuel burnup.

10-1

Axial Blanket / (AB) Upper

/BI J- (UBI)

200 400 600 Exposure (Full Power Days)

Figure 8. 2,,3 Am buildup and burnup 1n 100 MWd/kg

187
In conclusion, then, it should prove less risky to do our calculations on a CDC

machine than on the IBM one though the need is not fully demonstrated. One might

especially be leary of a new IBM burnup code which has not yet stood the test of time

for adequate accuracy.

Theoretical uncertainties: Nuclear data

Only a few cross sections are known to within 2%. For most one is lucky to get 10% . For rare of f-the-beaten-path isotopes, much larger uncertainties may be expecte - Fortunately, a recent demand for decay heat calculations has added some 800 fission

product nuclei to the ENDF/B files, including n,y cross sections for the nuclides of

interest here.

Table X shows that a 1% error in cross section can lead to a greater than 1% error

in production of a fission product nuclide ( llt5Nd). However, usually the error is

reduced, not magnified.

Table X Sensitivity of the computed production of stable fission products used in correlation

techniques (in %) to an error of 1% in the yields and cross sections of the

contributing fission products

[ ltt5Nd] sensitivity to sensitivity to itributing

isotope FP FP cross FP FP cross yield section yield section

141 Ce 0.0038 0.0006

ltt2 Ce 0.0023 0.0020 0.0004 0.0004

142Nd 0.004 0.0006

llt3 Ce 0.149 0.0042

l*+3p r 0.0007 0.0004

14 %d 0.0290 0.0120

l ^Ce 0.113 0.0073 0.03 0.0033

^Nd 0.0003 0.26 0.0695

145Nd 0.726 1.48 0.409 0.483

li+6 Nd 0.514 0.0362

Analysis of Existing Comparisons of Theoretical Predictions to Experiment

Virtually no comparisons have been made for fission product isotopes. Numerous com' parisons have been made for actinides, especially for the more conventional isotopes of

uranium and plutonium. For these agreement was quite good. For more exotic actinides,

agreement is not good; one suspects that the nuclear data are inadequate.

Table XI summarizes some of the theory-to-measurement comparisons found. A wide ra of errors is seen. For ratios involving isotopes 239 through 241 of Pu, agreement is

generally better than 10%, with some success indicated to the order of 1%; 21+2 Pu deviati

are generally closer to 10%.

In Figures 9 through 10, errors are compounded: e.g.,

21+0 Pu/ 239Pu will have errors from both isotopes, in contrast to 21+0 Pu/Pu for which the error in Pu should be much

smaller. The largest errors in this series occur in Figure 10 which involves high-error

21+2Pu end of the chain) and 2l+0Pu (intermediate chain member). Error bands of + 10%

indicated to aid the eye of the reader.

Figure 11 shows similar information for the Calder Hall reactor and U.K. calculatic

Two calculational methods are included. Agreement seems better for 2ltl Pu than for 21* 0 Pi This emphasizes the compensating errors which must occur for each nuclide in the chain

238u ^ 239Pu _> 2i+0p u + 2kl ?u 2 lt2p u as well as for Pu = E ip u . This limits the usefulr

of this information to our concern of calculating f ission^product isotopes.

188
Table XI Summary of Some Theory-to-Measurement Comparison Errors (%)

Calc of N-Reactor at Rich, Wash. VAK (BWR, 1971)

atio HEDL-1 HEDL-2 ARHCO-1 ARHCO-2 RMS Signed I 239 /Pu +0.1 +0.26 -0.1 +0.11 7.3 +4

240/Pu -1.03 -1.5 +0.1 -0.5 11.2 +4

- 241/Pu -0.52 -2.6 +1.9 -2.9 16.7 -17

242/Pu +1. +2.3 +10.5 +9.8 17.1 -7

Yankee

cores

4-6

+4.5

+1.5

+3.1

-12-

VAK (BWR)

1975 Calc,

Msmt . by

WAK

+1.1

-13.8

-39.9

+5.8

KRB (BWR)

1975 Calc,

Msmt . by

Eurochemic

+0.1

+2.6

+3.2

-12.4

Saxton-II

(20 GWD/MTM)

W

+0.86

+2.2

+3.6

+0.4

Saxton III (50 GWD/MTM)

Calc. by W (1973)

Range

-1.85 to 17.5

-15.0 to -3.1

0.8 to 9.8

-8.8 to 17.4

Avg.

+4.0

-8.4

+6.0

-13.8

— Pu/U

(w/0)

riot error band

2<,2pu/2-Qpu (weight ratio)

0.2 0.4 figure 9. Pu/U vs ' k0 Pu/ ?5, Pu; comparison with experimental results

from Trlno reactor (fuel enrichment 3.13 w/o). Figure 10. Pu/U vs lHZPu/3h0P\t; comparison with experimental results

frotr Tino reactor (fuel enrichment 3.13 w/o,.

1

!

I

MIMS 0 w1t« rev. graphite

scattering law A

'"-Pu

ARGOSY

UIMS 0 y 2 x 2 --Pu KEXPER1ME iTAL

ARGOSY

-

/ UIMS 0

y <~^=l 2900

n

3000 400

1 j

1 500

4

)MUd/ts

'"U/Total U (5)

Figure 11. Fraction of z -°Pii and 2 "Pu 1n total plutonlum.

189
Table XII is significantly different: here one makes comparison for absolute concei \ trations. As expected for absolute values, agreement is worse for even these very standai

nuclides

.

Table XII Comparison of calculation to isotopic data measured at the

reprocessing plant on groups of PWR fuel assemblies

Fuel Assemblies Depletions (kg/t(U)) Production (kg/t(U))

0 3 £ 236 D

Z40Pu Pu fissile

C"E r, v C-E„ Dev.%

E Dev . a> E Dev.%

E Dev.%

E 16-G5-E8 +3.0 +2.6 +6.8 +4.29

E4-15-H4 +4.3 +3.0 +9.5 +6.1

17-G10-F8 +2.9 +1.9 +4.8 +1.00

D5-D8-F4 +4.1 +3.1 +4.8 +1.8

H6-F7-F6 +2.3 +0.9 -1.0 -2.37

GP-F5-J6 +2.4 -0.05 +7.2 +4.14

E9-G4-E5 +3.9 +2.6 +5.5 +2.94

F9-I8 +3.9 +2.8 +5.8 +2.35

E11-K6-C8 +4.6 +0.2 -9.0 +4.1

E8-I10-F11 +3.8 -4.5 +8.7 +4.12

G11-D4 +3.3 +0.6 +8.1 +4.66

to Figures 12 through 15 show the ability of calculated ratios to follow variations due

(1) initial fuel enrichment,

(2) BWR void fraction, and

(3) moderator ratio.

190
O.l

236U/ 238 U (^efqh ratio)

/// /

1

i / V . ANKEE ROWE

(MR « 1.286)

TRINO

(MR 1.88)

enrichment (w/o)

2 3 4 5 Figure 14. 2?4 U accumulation/ 23 ^ depletion ratio; comparison with experimental results from Trlno and Yankee Rowe reactors.

0.'

0

I Pu/U (weight ratio)

1—

ANKEE ROWE

MR = 1.286)

TRINO

(MR = 1.88)

;nrichment (w/o)

2 3 4 5 Figure 15. Pu/U to 235U depletion ratio; comparison

with experimental results from Trlno and Yankee Rowe

reactors.

that D5 means depletion of 235U. Dresden and Humboldt Bay are BWRs; Trino and Yankee

are PWRS . Due to sophisticated codes like CONDOR-3, they are able to reproduce some

ie sensitivities, but not all. The same sort of success is shown (Figures 16 and 17)

ying to calculate spatial heterogeneity effects (pins near boundaries, control rods,

•ecial channels) . Codes like CONDOR-3 do show that they can reproduce some of the

tivity but they are not exactly right on the money.

25 5 U Oepletton (Kg/t-(U))

Figure 16. Fissile plutonlum production vs 235U

T— —tr

235 U Depletion (kg/t (UH

Figure 17. 2 *°Pu production vs 235U depletlo

1n different spectra.

191
Table XIII shows the marked improvement one can achieve on perturbed (corner) rods

increasing the spatial and group mesh. Note that for the higher isotopes agreement get worse with better techniques - indicating that some, but not all, of the compensating

errors are being removed .

Table XIII

Comparison of several calculations to several experiments: (C-E)/E (%)

(Coarse) BUKNY Code (Fine) BURSQUID Code

Corner Rod Stand. Rod Corner Rod Stands Rod

BURN-UP (Cs) 1.82(-) 1.17(+) 0.60 1.60(+)

[u-235

(Nd) 3.25(-) 0.49 1.35(-)

6.04(-) 1.15C-) 1.24 1.42(-)

[U-236" 2.65(-) 2.66(-) 3.60(-) 2.2K-)

[U-238 0.10(+) 0.05(+) 0.04(+) 0.02(+)

[Pu-239[ 16.5 (-) 6.4K-) 2.10(+) 1.17C+)

[Pu-240 5.07(+) 2.00(+) 1.28(+) 1.09 (+)

CPu-24l' 6.35 2.27 4.43 3.43(+)

fPu-242 6.90 1.98 9.22(+) 8.35(+)

1 iedi

11:

Figure 18 tends to show simply that its much harder to predict for a sample from

end of a rod vs. from its middle.

PERCENT DEVIATION

Axial Asymptotic Intermediate Perturbed

Location Rods Rods Rods

Net grams of fissile

isotope destroyed per MWd

Fuel Assembly G-7

Fuel Assembly H-10

Indicates Karlsruhe results

Figure 18. Relative predicted-measured errors on net grams

of fissile isotopes destroyed per MWd and on local burnup.

In summary this section shows that success in accurate prediction of measurements 1 not come easily. It will necessitate sophisticated techniques to eliminate calculation

errors. The trick will be to get advanced techniques which are well suited to the task

and avoid any expensive overkill. Once that is done it becomes important to choose

nuclides for which nuclear data uncertainties are small. The use of sensitivity calculi

tions with the ENDF/B covariance files may be necessary for full understanding. Thus tl job can be done, but it requires considerable care, patience, and money.

Decay Chains of Interest and Their Requirements Upon Computer Codes

Existing Information on Transmutation Chains and Associated Nuclear Data

For the moment it is convenient to reclassify mutants as follows:

1. actinide mutants

2. radioactive fission products, generally measured by non-destructive gamma

spectrometry.

3. stable and neo-stable fission products, generally measured by chemical analyse:

and mass spectrometry.

4. cladding mutants

192

Rial
stigation of production of the main actinides has been motivated by concern over fuel

ltory, hazards, and economics. Thus decay chains and nuclear data necessary for

retical calculations have been developed. Cross sections and decay constants for the

exotic actinides are being added to ENDF/B-5 (available in 1978). Similar data for

radioactive fission products have been built into zero-dimensional codes like ORIGEN

BINDER whose purpose is to calculate decay heat. Personal communication with the

ent authors indicates that the stable and neo-stable fission products are also adealy represented and the specific chains of interest here are included. ENDF/B-IV

Lded the first cut at yields, decay constants, and cross sections for 800"1" fission

act nuclides. ENDF/B-V will include numerous improvements. Little data development

r.he cladding mutants has occurred due to lack of interest.

Thus for the chemical analysis data of interest here, the nuclear data and transmutachains needed for prediction tend to be well developed though still undergoing some

5es. The data libraries existing in a few codes are mostly for LWRs, especially PWRs.

rally they are in the form of four-group (or less) constants, having been previously

~ed over a typical PWR spectrum. With that, the zero-dimensional codes tend to predict

r distributions within 5% and heavier nuclei within 20% in asymptotic spectra. FP

irisons have not been made.

Unfortunately, the incorporation of these chains and associated data into codes like

ER and ORIGEN does not in any way facilitate incorporation into other burnup codes,

lata are still described in only one place: ENDF/B and limited accompanying

aentation.

On Looping

Chain steps are very dependent upon the product of o and <J>. Given the fact that we

usually considering fuels which have been exposed to their full burnup potential, then

suspects the possibility of some chain looping where a chain path can lead to a previous

sr. Most codes cannot handle this. This may occur in both FBR's and LWR's: In the former

high; for the latter, a is. Still the problem is known to be more serious for actinide

is than for fission product ones. Decay Chains for Use in Predicting Fission Products of Prime Interest

Figures 19 through 23 show five chains deduced in this work for the isotopes of

rest. The data displayed on the figures represents typical values used to determine

prevalent paths. The latest data are available in the ENDF/B library (current

Lon) or from EG&G's Nuclear Physics Branch (current evaluators)

.

Cunulatlve

A-Cha1n

Yield 5.8

(»)

6.1 6.3 5.0 4.2 1.8 1.0 0.39

"Mo (38)

" J Tc

Cumulative u v^Chaln Yield

2.13 x 10 5 t, years cross section (b)

99^^(l0lVLM'm>-i-^ll2>^rt03) (]04)-^M!05) (106; 4.44 1 369

Direct

Yield Only

Figure 19. Decay chains for pertinent Ru Isotopes.

193
Cumulative

A-Chain

Yield (%) 42 7.7 6.6 6.3 6.3

"Cs

9.17

hr

30 „ 140 _ ,lr 9 133

10 > 1 34 TM > 135

I 2.05 I y

137}

Figure 20. Decay chains for pertinent Cs isotopes.

Cumulative

A-Cha1n

Yield [%) 6.3 5.9 5.9 6.0 5.5

"Ba (138)-

"La

5»Ce

140

(139)

9 - 2 » 140

33.0 284.4

Figure 21. Decay chains for pertinent Ce isotopes.

Cumulative

A-Cha1n

Yield (I)

6 3 5.9 5.9 6.0 5.5 3.9 3.0 2.3 1.7 1.1 0.65

141

2.53 133.0

d I |» "4-142 143

I 19.16 I 13.5 1 h Id

142 l^43M^44-l*^^^i4»^^-@i4-@) @J

T 10-99 |1.73

Figure 22. Decay chains for pertinent Nd Isotopes.

Cumulative

A-Cha1n

Yield (%)

„ 65 0 42 0 27 0 i6 0-075 0.033 0.014

''Eu

Figure 23. Decay chains for pertinent Eu Isotopes.

194
In deducing the chain paths, neutron absorption is in competition with decay. For

sorption to be significant (>1%) it is necessary that

> _L. 0-693

C 100 4> • t-|/2

rr , ., , rt14 -2 -1

Taking <j> - 10 n cm sec one gets

> 2.2 = 802. = 19,250

fc l/2 (y ears ) t l/2 ( days) t 1 / 2 (hrs)

Examination of these chains leads to some first-order conclusions:

(1) much cross section information is relatively unknown.

(2) most of the isotopes of interest have time constants (from decay

or neutron absorption) of the order of years - of little value here.

(3) the production of some isotopes stems from more than one path controlled

by cross sections. The accompanying data uncertainty will probably limit

seriously their usefulness. For history deductive purposes the best ratios

should involve a neo-stable isotope in the denominator and ones with a time constant of the order ^ 2 to 50 weeks in the numerators. For the

denominator the element may also suffice as it is relatively stable.

For Ru, 103Ru/Ru and 106Ru/Ru are of interest. Isotopes 100, 101, 102, or their

a might also suffice in the denominator. For Cs, beware the problem of Xe precursor

'ration. The best Cs is therefore 135 Cs whose Xe precursor has a half-life of only

L2 hour and 137Cs (precursor t]_/2 = 3.84m). However 137Cs also is produced through the

= 136 chain where the Xe is stable, and 137Cs migration is already known to be a

Dblem.

For Ce, 11+1 Ce and lttl+ Ce have attractive time constants. llt2 Ce should make a good

r erence. For Nd, only 11+7Nd comes close to having an acceptable time constant. For

a number of isotopes look interesting, especially because of large cross sections

sotopes 151, 153, 155). Their true value would take more extensive investigation to :ermine due to the complexity of production and destruction paths.

Analysis of Code Features Under Consideration

On Multi- dimensional Capability

It was earlier shown that isotope production is sensitive to regions of different

'ichment. In a comparison of calculation and experiment, one saw a difference in con- izations between asymptotic spectra (near the center of a region) , and perturbed spectra

:ar a boundary or control rod) . The two-dimensional methods employed were able to

.culate most of these differences. Although one might be able to find mutant ratios

.ch are insensitive to these effects, they would then also be insensitive to phenomena of

.erest to this study. Thus, although zero-dimensional calculations may suitably cal-

.ate some isotope correlation ratios, those of interest to this work will require

.tiple dimensions. The question now becomes what will suffice? We certainly can't

'ord 3D transport theory recalculations at each burnup step (the extreme)

.

Table XIV indicates a fine point: for PWR's adequate representation of radial

:erogeneities appears to be more important than full axial representation.

Table XIV

Regionwise Burn-ups Evaluated With Different Methods (MWd/t(U))

Inferred from r,z-C0ND0R-3 x,y-C0ND0R-3

Region in-core instrum. calculation calculation

Inner 12 933 12 528 12 985

Intermediate 13 192 13 183 13 104

Outer 8 648 9 068 8 675

195
Tools Needed for PWR and BWR Analysis

There are important differences between PWR and BWR which significantly affect the

prediction of transmutations. In the modern PWR, reactivity is primarily controlled by

dissolved B which is uniformly distributed in the moderator. The reactor is designed to avoid significant boiling in the moderator such that the variation in moderator-to-fuel \ volume ratio is small axially and radially. Dissolved B control and variable radial

fuel enrichment minimizes radial and axial peaking. These features make the calculation

of element averaged concentrations for a PWR relatively straight forward. In a BWR the

moderator-to-fuel ratio varies substantially in 3D. Also, rod insertion is used for

control.

A saving simplification in BWR analysis lies in the high degree of standardized

similarity between BWR's, especially the later standardized designs (Table VIII). A

large uncertainty remains in the control rod patterns. If some nominal control configuration could be standardized, the BWR's could be characterized by 3 or 4 FLARE-ISOCHECK

calculations

.

Outline of a Typical Complete Set of Burnup Prediction Methods

The techniques for detailed burnup evaluation are essentially based on:

(a) burnup dependent cross section library calculations whose fundamental tools ar

a multigroup treatment of slowing-down; a representation of the resonance

capture; a lattice-thermalization law together with a fine-group representatio

of the disadvantage factors; and solution of the time-dependent equations for

variable isotopes using cross-sections collapsed into four groups.

(b) power- and burn-up-dependent corrections to cross-sections. Mainly the 238U

cross-section dependence on power shapes through life-time is evaluated.

(c) power- and burn-up shapes evaluation during life-time using a coarse or detailed representation of the core. The main codes for this purpose are:

a one-dimensional diffusion-depletion code (radial or axial) including

special features like spatial Doppler variation; spatial moderator temperature

variation; burn-up-dependent cross-section libraries; criticality searches; am two-dimensional diffusion-depletion codes with the same features as the one- dimensional above, together with composition-and burn-up-dependent cross-sectii

libraries

.

Sometimes an iteration procedure between spectrum and local flux evaluation is

applied in order to define the proper spectrum for each composition. When survey calculations or experimental data are available, the composition- and burn- up-dependent

library evaluation is performed using the critical spectrum in conditions very near to the real ones (local power, poisoning and leakage), otherwise spatial flux distributions

are evaluated using average cross-section libraries and the appropriate spectrum for eacl composition is calculated again.

A simplified synthesis technique is often used to estimate three-dimensional power

and burn-up shapes. The model consists of the detailed x, y-representations of the

cross-sections of the core, using local poisoning (for the representation of partial

control rod insertion) corresponding to the reactivity effect of the absorbers, diluted

poison as required, composition-dependent axial leakage (obtained through survey

calculations), burn-up- and composition-dependent cross-section libraries. The point-wif

burn-up evaluation is carried out, and the compression of the results by compositions or' regions during burn-up gives the inputs (local power, fast-to-thermal flux ratio, leakagt

poisoning) for the axial representation in the regions of interest. The axial burn-up

calculation gives all the burn-up dependent characteristics at a given point or region

in the core.

A more sophisticated model is used when perturbations due to control rods are predominant during burn-up. The model includes an x, y-calculation as previously and, when

a cylindrical representation is possible, an r-z or radial calculation during burn-up,

or a wide-mesh three-dimensional burn-up evaluation.

196
Extensive use of the mixed number density model is made. Problems related to burnable

ons behavior are treated with special routines for burnup of poison spheres and

; nders

.

Deduced Code Requirements and Desirable Features

Required Code Features

1. 1-D for PWR's, LMFBR's

3-D (FLARE) for BWR's

2. Multiple groups for LMFBR's

3. Adequate numerical ability

4. Burnup plus HARMONY-type representation of burnup-dependent cross sections,

or ability to recalculate fluxes periodically.

5. Available or easily- imp lementable nuclear data library

6. Error propagation stability with time step

7. Feasible cost of running

Highly Desirable Code Features

1. Able to input different fission yields for 234 > 6 » 8 U and 239-lt2Pu.

2. Calculation of BD in fuel elements next to control rods, water reflectors,

or other large-flux change regions.

3. Could use zero-dimensional capability for exploratory surveys such as effect of excursions

effect of reduced power operation

- which fissiles contribute the most?

4. Available on a CDC computer

5. Ability to read in chains and associated data

Other Desirable Features

1. Checked out on IBM as well as CDC (for possible use on IBM if numerical error is

not a problem to chain members of interest)

.

2. Pre-fabricated modern chain library, ready to go.

3. 3D inventorying.

4. Multi-dimensional fuel movement and inventory representation.

5. Looping ability of burnup chains.

6. Local experience.

Recommended Code Systems

From all of the information above, it appears that a code system which includes

i- dimensional neutronics calculations is needed for the purposes defined here. A good

.pie is Ispra's system built around C0ND0R-3. An American equivalent of C0ND0R-3

-d be PDQ-7 or -8. The use of HARMONY for burnup-dependent cross sections with PDQ

Is a powerful tool. Indeed, this package has been used for many light water

tor calculations.

What this package might lack is

(1) routines or codes to handle fuel management

(2) automation of intermediate steps

I (3) a cheap way to do BWR calculations which require an extra spatial dimension

(possibly a synthesis option in some PDQ versions may help)

.

Another attractive solution is the new EPRI-sponsored ARMP package. Designed to be

by public utility engineers it

(1) removes many headaches of data handling,

(2) is designed for light water reactor calculations,

(3) has a FLARE (GE BU'R code) - type subroutine to allow economical calculation of

a third dimension (for BWE.'s especially), and

(4) includes fuel management routines

.

e Indeed, this package was designed for a purpose which highly overlaps that declared

It presently uses a cross section library of ENDF/B caliber, but a new one from

197
ENDF/B-4 is imminent. Further, it is logical to expect that EPRI will continue to keep

both methods and data current for a long time.

To calculate the fission product nuclei of interest will still require some chain

formulation read-in along with appropriate nuclear data processing. However, this will

be true of all codes except for ORIGEN, RIBD, and CINDER which calculate everything under

the sun. Their inadequacy lies in their zero-dimensionality for neutronic calculations.

(Actually CINDER plays a small role in the ARMP package in preparing burnup-dependent

cross sections for HARMONY)

.

The ARMP system is proprietary to EPRI and NAI, a CDC subsidiary. It will be

available at all CDC utility computer service centers (at their computer rates) and at three ERDA/NRC computer installations: BNL, ANL (Chicago), and UC-Berkeley.

SUMMARY CONCLUSIONS

Sources of Information

Numerous individuals and labs have been active in fields related to this area of concern. A large number of spent fuel analyses have been made but it is a difficult

task to locate the results and a description of the reactor and its history, and then to develop an appropriate nuclear model. Much of the information does not appear to be in

the open literature. One major source would be the data bank at Ispra. A similar bank

may develop in the U.S.

The numerous empirical correlations could prove useful as benchmarks, but the

original data is to be preferred. Presently the empirical data would seem more useful

as indicators of the sensitivity of an isotope ratio to various reactor parameters.

Cursory survey of the large number of burnup codes and code systems indicates some basic trends. Comparisons of theoretical calculations with measurement are rather limited for fission product nuclei.

Workers at SGAE-Austria are among the very few active in the field of a backward

(deductive) calculations.

Problem Understanding: What Are the Mutant Parameters of Interest?

Interruptions to a reactor's operation may be detectable through analysis of spent

fuels with zero - or one-dimensional techniques. To sense where fuel substitution or diversion has occurred, appears to require multi-dimensional techniques. In the

situation where one knows (or wishes to assume) nothing about the origin or history of

a fuel element, it is important to perform visual and gamma spectrometric analyses. These

can help identify the type of element and determine important heterogeneities. Neutron

radiography is also a possibility. A logical sequence of analysis should be developed

leading to identification of reactor type and composition before attempting to resolve

detailed history. This is important so that the correct weighting spectra are used in

the latter effort.

In studying the spent fuel analyses one can use isotope/isotope ratios (Ax/^Y) and

isotopic compositions (^X/X) . The latter are more accurate from the measurement point of

view as the element and isotope should have the same chemical recovery efficiency.

Theoretical uncertainties also should be smaller for the denominator X (errors partially

cancel when summing up the isotopes of a given element) than for ^Y.

Most of the isotopes of declared interest in this study have not been measured or predicted in other studies. 137Cs and 11+8 Nd are the notable exceptions. Much valuable

information in the past has been extracted from the actinides. As nuclear cross sections

and decay data are well known for the more conventional nuclides and no fission product

yields are involved, their use in forward and backward calculations can not be ignored.

Also considerable experience including comparison of calculation with measurement has

been developed for them.

198
The work to date on backward calculations provides some valuable insights. A key idea

assume a pattern of history and parameterize it. One can then solve for the paras using the same number of isotope ratios. Use of a larger number of isotope ratios

overdetermine the problem: this should make up for the uncertainties in measurement

heory. It should also indicate, by difference in answers, the validity of the

ed model. Several models would be tried.

Judicious choice of isotope ratios is important: a scientific basis for logical

e appears possible. A theoretical understanding of some standard nuclide concentrai behaviors was developed. To maximize the information obtainable — to choose the

ts of interest scientifically rather than by trial and error — theoretical investi- n should be continued. Specifically one would like to develop permutational relation-

! which could approximate impulse functions of varying widths. Approximation of

mathematical functions might also be possible and of interest.

Problem Understanding: Considerations in Comparing Theory to Measurement

It would be well to attempt to quantify the uncertainties due to (1) isotope migration

m a rod and (2) viable chemical recovery efficiencies. Elements with similar high

iencies and low variances would be particularly desirable for forming ratios. The

i ted variation in measurement uncertainty among isotopes of the same element should

plained (for better understanding of the problem)

.

One tends to prefer the CDC computer for burnup calculations though an IBM code

uble precision may be adequate; in any case we suspect a CDC code may be more

ient here, whichever computer code is chosen should be well documented and supported,

which have stood the test of time tend to insure against errors stemming from

ter word size or program logic.

Errors in nuclear data may be expected to play a significant role here. Sensitivity

sis shows that uncertainties in yields are every bit as important as those in cross

1 ons. Therefore, it behooves one to use the latest ENDF/B data, be aware of the

tainties and eventually do sensitivity studies.

Evidence exists that key reactor lattice parameters sufficiently affect the neutron

I rum such that isotope concentrations are significantly affected. Furthermore, (at

) American PWR's and BWR's each exhibit a narrow range of lattice parameters in their

i lass: this will facilitate distinguishing between the two. Direct evidence exists

lone can detect a long stand-still of a reactor by means of the nuclide compositions.

In seeking out mutant ratios to be studies, one should bear in mind that isotopic

sitions (e.g., ^Z/X) appear to be inherently more accurate than nuclide ratios

JY) . As both measurement and theoretical uncertainties exist, it will be necessary

erdetermine the analyses. Also, any one nuclide ratio will have a blind spot or

i sensitivity to some reactor or history characteristic. This means using at least

ore ratio than model parameter sought. A variety of time constants should be

I ded among the nuclide ratios.

) Theoretical research should help to develop a scientific method for choosing nuclide

t s best suited for each question asked. To this end it would be well to set forth

1 evelop a proper set of questions which is to be asked of the spent fuel cycle

1 ses

.

Information from Existing Comparisons of Theoretical Prediction to Measurement

Virtually no comparisons were found for fission products; only for actinides. In

dering the conventional isotopes of plutonium, agreement varies. However in a

ell, when reported in ratio to total Pu, isotopes 239-241 look pretty good (^ few

or) while 242 exhibits K 10% error. This is not surprising as the 242 cross sections

he least well known and its presence at the end of the chain has the effect of

porating the errors of all its precursors as well.

199
Some significant evidence demonstrates that reasonably good agreement for higher

actinides is entirely fortuitous and due to compensating errors: improvement in the

calculational technique makes agreement worse.

Existing sophisticated calculational techniques are able to reproduce a good porti

but not all, of the sensitivities of isotope ratios to spatial and spectral perturbation

The point is thus emphasized that one must carefully use the best techniques if the

required precision is to be obtained. Judicious choice of the isotope ratios will also

important to avoid those nuclides with large associated data uncertainties.

Decay Chains of Interest and Their Requirements Upon Computer Codes

The decay chains for the fission products of declared interest here were deduced

and examined. In the context of other sections of this report, it was found that most

isotopes would give the same information, and therefore be redundant. Others were

clearly unique. Thus it is very important that the characteristics desired of the

isotope ratios be further investigated, classified and applied to the set of isotopes

of possible interest to the situation to be analyzed.

The resultant set must also satisfy data requirements as indicated above. Sensitivity analysis should prove to be valuable: first, sensitivity of isotope concentratir

to the different chain paths so as to define its appropriate class and time constant;

second, sensitivity of isotope concentration to the uncertainties of the underlying

nuclear data to see if it will be useful.

Analysis of Code Features Under Consideration; Requirements and Desirable Features

Multidimensional neutron calculations seem to be a requirement. An appreciable

number of codes will be required for cross-section processing, 2D neutronic calculation,

3D coarse mesh neutronic calculation, burnup calculation, and inventorying with or withe

fuel shuffling. Linked systems of codes are available which can significantly

facilitate the whole process.

Recommended Code Systems

Items of major significance observed here are:

(1) limitations of zero-dimensional codes

(2) excellent suitability of the EPRI ARMP package to this problem.

(3) the large number of nodal method codes coming into existance. These allow

one to perform a 3D calculation with suitable approximations and reasonable

cost (compared to a 3D finite difference calculation).

The recommended calculational system would be one which includes:

(1) 2D neutronics (like C0ND0R-3 or PDQ-7) , and

(2) coarse mesh, nodal code for 3D (possibly 3D synthesis will suffice)

By far the best system seems to be the EPRI-ARMP: it has virtually everything

needed, it is highly automated and human engineered, and doesn't overkill the problem

(minimum cost to run). Should access to this semi-proprietary system be infeasible,

other existing capabilities could be used: they're just not presently automated, and

are less efficient. Modularized computing capabilities would facilitate assembly of a suitable code system.
In-Situ Transuranium Element Measurement Technique

for Wastes Associated with Power Reactor Fuels

K. K. NIELS ON. R. L. BRODZIMSKI and N. A. WOGMAN

Battelle, Pacific Northwest Laboratory, Richland, Washington

ABSTRACT

A planar, 19 cm2 intrinsic germanium detector has been used for insitu analysis of plutonium and americium in contaminated laboratories and

buildings. Detection limits depend on local background activity, but in

typical surface measurements for decontamination work are about 0.005

nCi/cm2 for ZhlAm (59-5 keV) and 0.5 nCi/cm2 for 239 Pu (lj .2 keV) . Specific analyses of 2 3 8 Pu, 2 3 9 Pu, 21t0 Pu, and 21tl Pu are also possible using

various gamma- rays . Attenuations equivalent to 10 cm of concrete can be

tolerated for high levels of 239 Pu and 2klAm.

KEYWORDS: In-situ analysis, americium, plutonium, detection limits, germanium detector, x-rays, gamma-rays

INTRODUCTION

Increasing precautions being taken in nuclear materials handling and management have

red the development of instrumentation for rapid, sensitive analyses of various transc waste elements from both power and breeder reactors. Such instrumentation is red to analyze residual solids in high and low level waste streams and on contaminated

ment, and to provide accountability. Since gross beta, alpha, or even neutron flux

jirements frequently give insufficient information about the concentrations of transuran-

.ements , the analytical capabilities of x-ray and low energy gamma-ray spectrometry have

investigated. Transuranic elements which are readily identified in this manner include

:, 239 Pu, 21tl Pu, and 21,1 Am.

This paper describes the first field tests of an intrinsic germanium detector for inanalysis of low energy photons from 239 Pu and 21flAm in fissile material storage and

•atory buildings and in a building undergoing decontamination. The sensitivities, de- on limits, and other information such as the depth of the activity in an absorbing

Sm are given.

METHODS

iA planar 19 cm2 intrinsic germanium detector, attached to its cryostat by an adjustable

.ing (Princeton Gamma-Tech, Princeton, NJ) permitting 180° adjustment from up-looking to looking or at intermediate positions, was used for all x-ray and gamma-ray analyses,

.etector preamp was coupled by a 30 m cable to the amplifier. Pulses were stored in a ble multichannel analyzer which was interfaced to a portable cassette recorder (Model

Techtran Industries, Inc., Rochester, NY) for storage of spectral data. Nominal der resolution at 59-5 keV was 0.68 keV FWHM with this apparatus. Unfortunately, just

• to the field tests a defect developed which degraded the system resolution to 1.0k keV.

jiovable annular lead collimator (1.27 cm thick by 10.2 cm long) was placed around the

):tor head to reduce background and restrict the detector's field of view.

Laboratory calibrations were accomplished with thin sources of 21tl Am, 23 8 Pu, 2 3 9 Pu, and

i, counted at various distances from the detector for geometry assessment and through

>us absorbers for determination of attenuation. The attenuation measurements for 2kl Am

used to estimate linear coefficients relating L x-ray peak areas and relative attenua-

^actors to the observed 26/60 keV peak area ratio. This permitted attenuation correc-

; for the z '* 1 Am measurements and interference correction of the plutonium L x-ray peak

Once corrected for interference by the partially resolved L x-rays from 241 Am, the L

's from plutonium decay could be similarly corrected for attenuation using the LR /L peak

ratio . " Y

201
Besides correction for photon absorption, the peak ratios were also used with appro

mate attenuation coefficients to estimate the thickness of any absorbing material over t transuranic element source, or in the case of a uniformly distributed source, the mean as of the activity. Attenuation coefficients required for these depth estimates were measu

in the laboratory using various materials such as concrete, wood, sand, steel, lead, luc

glass, and floor tiles. Additional sources consisting of several transuranic species mi' in coarse sand or finely packed dunite were also analyzed to verify the attenuation calci tions . Details of the calibration and calculation methods have been described elsewhere

Analysis of contaminated surfaces consisted of positioning the detector, acquiring

for 1-10 min, recording the spectrum on a magnetic tape cassette, and moving to a new lo

tion. Analyses both with and without the lead collimator were done to evaluate its use effectiveness. Although prominent peaks such as the L x-rays and the 26 and 60 keV gamm

rays were immediately observed, detailed analysis of the data for plutonium activity and

depth was done following transfer of the spectra from the cassettes into a PDP-15 comput

itJ'

RESULTS

The first of three field experiments was an attempt to measure surface contaminatio

a concrete room containing large arrays of steel drums of fissile material. Background

this room created ^60% dead time in the unshielded detector and prevented observation of

photopeaks below the intense 60 keV 2 Am peak. Although the annular lead collimator re duced the dead time to ^30%, it was still not possible to measure low level local surfac

contamination

.

The second experiment consisted of a series of measurements near a pair of contamin

hoods in a chemical separations laboratory area. Measurements were made at three locati

(60 cm intervals) on the floor in front of the hoods and also looking horizontally into

hoods . The contaminated portion of the room was isolated by a thin plastic cover which

not significantly attenuate the observed peaks. Spectra collected at a location on the

both with and without the lead collimator, are illustrated in Figure 1 (Table I, Locatio

oil ii%

is,

ictei

10'

10'

10

10'

10

COMBINED U AND Np X-RAYS FROM

24lAm AND Pu

. I 1 . p ' - 241

DETECTOR: 19 cm 2 PLANAR Ge

LI VET I ME: 300 sec. SOURCE: FLOOR BY CONTAMINATED HOOD 2216-5 SOURCE-DETECTOR D I STANCE: ~ 10 cm

COLL I MATED SURFACE AREA: ~ 66 cm 2

NO COLLIMATOR

LEAD COLLIMATOR

238, 239, 241

241

Pu,

Am,

99

241 Pu

98

239, 241

95

Pu

241 Pu,

241

239. 241

Am

101

239, 241 Pu

111

I

100 200 300 400 500 CHANNEL

600 700 800 900

Figure 1. Low Energy Photon Spectrum from a Contaminated Floor Area

1 K. K. Nielson, C. W. Thomas, N. A. Wogman and R. L. Brodzinski , "Development of a Plutoi

Americium Monitor for In-Situ Soil Surface and Pond Bottom Assay," Nucl . Instr. and Metl 138, 227-237 (1976). 202
As can "be seen, the x-ray peak areas are not appreciably reduced "by collimation, which

iicates that the surface contamination of this particular area is local and confined to viewing area of ^6 cm2

. The elimination of the higher energy plutonium gamma-rays

: -120 keV) and attenuation of the 59-5 keV 21fl Am peak by the collimator indicates detecn of these higher energy photons from beyond the 66 cm

2 viewing area, most likely through

sides and top of the unshielded detector head. The observed concentrations and depths

activity of the contaminated floor areas are summarized in Table I as calculated by the

k ratio method, 1 using L x-rays for plutonium determination and the 26 and 60 keV peaks

21* l Am determination.

Sampling Location

TABLE I

Floor Activity (nCi/cm2 )^

1 2

2 3 9

l Am

Pu

0.083 ± 0.013

1.2

Remote^1

0.1+8 ± 0.05

1.9 ± O.k

2.8 mm

1.9 ± 0.1

29 ± h

; Mean activity depth 2.8 mm

Statistical uncertainties only.

^No peaks were observed below 60 keV indicating activity was deep or, more likely,

simply background.

The hoods themselves were examined at a distance of ^25 cm, yielding a collimated viewarea of ^k00 cm

2

. The results of the hood analyses are listed in Table II. For Hood A,

of the lead collimator greatly affected the observed spectrum, as illustrated in Figure

nd Table II. In this case, the entire spectrum was attenuated, including the x-ray

ks, indicating the activity to be spread over a large area which was geometrically re- icted by the collimator.

100 200 300 400 500 600 CHANNEL

700 900 1000

Figure 2. Low Energy Photon Spectrum of a Contaminated Laboratory Hood

203
The minimal reduction in photopeak intensities made by the collimator in analyzing Hood B indicates the activity was almost entirely confined to the a.1+00 cm2 viewing area.

TABLE II

Hood Activity (nCi/cm2

)

Hood A Hood B Without With

Collimator Collimator

Without With

Collimator Collimator

2 39 24 1 Pu

(95 keV)

Pu

(L x-rays

Am

(60 keV)

20T00

5811

590

5^50

318

1+1+

33900

1090

730

33^00

1+1+6

690

An additional spectrum collected from Hood B at a lower gain is illustrated in Figure

3 and includes a variety of higher-energy gamma-ray peaks from 239 Pu, 241 Pu, and 237U

( 241 Pu). These peaks extend the depth determination capability for 239 Pu by a factor of 1C and could potentially permit limited attenuation or depth profile modeling.

238, 239. 241

DETECTOR: 19 cm PLANAR Ge

LI VET I ME: 100 sec. DEAD TIME 20% SOURCE: CONTAMINATED HOOD, 2216-5 SOURCE-DETECTOR DISTANCE: -25 cm

COLL I MATED SURFACE AREA: ~ 400 cm 2

239 D

Pu,

241 239 q '^Am p u 239 D

374 p u 413

*«kir

100 200 300 400 500 600 CHANNEL

700 800 900 100

Figure 3. Gamma-Ray Spectrum of a Contaminated Hood Using O.i+7 keV/Channel Energy Scale

The third experiment was conducted inside a concrete cell which was undergoing extensive decontamination efforts including paint removal and, in some cases, removal of surface

layers of concrete. Two areas of suspected residual contamination were found to contain

204
023 ± 0.002 and 0.022 ± 0.002 nCi/cm2 of ZhlAm. Although plutonium L x-ray peaks were

bserved, the corresponding activity was below the 3o detection limit of .1+6 nCi/cm2. he 21tl Am 30" detection limits were 0.001+ nCi/cm2 for these ten-minute counts. In the ab- ence of x-ray interference from 21tl Am, the 2 3 9 Pu detection limits would drop to ^.13

jCi/cm2.

DISCUSSION

The in-situ detection limits of the field system described here are similar to those

istimated earlier 1 based on a smaller laboratory germanium detector housed in a lead cave. tree sigma detection limits for ten-minute counts have been computed from the background

ntensities of the in-situ spectra and are plotted in Figure k as a function of concrete

.bsorber thickness to illustrate the relative depth of penetration of the various photons

.

,he detection limits were computed as

Detection Limit = 3V2B" exp(uX)

,

S here S is the sensitivity (counts per nCi/cm2 ), B is the ten-minute background intensity

"or the peak integration region, u is the linear attenuation coefficient for the given

iihoton, and X is the thickness of the absorbing medium. As shown, 2ltl Am is readily detec-

•ed in the 0.005-0.1 nCi/cm2 activity range at depths to ^1+ cm, while plutonium detection

•equires activities 'VLOO times greater for surface counting and VL000 times greater at

Lepths of a few centimeters

.

The surface detection limits depicted in Figure k for the 59-5, 17-7, 17.2, and 26. k

i:eV peaks were estimated from the spectral backgrounds observed in the concrete cell experiment. The remainder of the detection limits for transuranic element photons are based on

iackground intensities observed in analyzing the contaminated hoods in order to more realissically assess the practical limits of plutonium detection. Since 21fl Am is generally found

In any plutonium contaminated area, the conservative plutonium detection limits given in

r igure h are more appropriate. Plutonium detection limits would be much lower than indica-

;ed by Figure h if computed on an interference- free basis. For example, in the concrete

:ell where the 21|1 Am activity was low, background intensities were lower at 50, 100, 200,

,|iiid 1+00 keV by 0.055, 0.012, 0.057, and 0.20, respectively. Detection limits for 21*°Pu and

Z38 Pu were calculated from relative intensity data 2 since significant peaks were not observed for these isotopes in this weapon's grade material. Both of these isotopes are much

no re abundant in high-burnup power reactor fuels .

Although the detector system used for these measurements was not operating with optimum

resolution, its efficiency and detection limit capabilities were not greatly affected. Although improved resolution would reduce interferences in the L x-ray peaks and also in the

55-117 keV region, the overlaps would still not be well resolved and analysis of these peaks

[Bust include some form of interference corrections . These are especially important for the

four peaks in the 95-101+ keV region which are potentially composed of up to ll+ x-ray and

gamma-ray peaks

.

3

Continued development of this in-situ analysis method for 241 Am and plutonium will inlude modification of the lead collimator to provide better definition of the surface area

oeing counted. In addition, quantitative measurements will be further aided by computer

analysis of the spectra for rapid activity and depth estimates . Semi-quantitative or qualitative estimates of the nuclide activities and depths can also be made in the field using

'analyzer-integrated peak regions or by visual inspection of the data as illustrated in

Figures 1 and 2. Approximate on-the-spot assessments of the quantities and locations of

specific nuclides are frequently adequate for decontamination work. However, in the areas

of nuclear safeguards or decommissioning of nuclear facilities , the ability to analyze

specific transuranic nuclides at the 0.005-1 nCi/cm2 level may be important.

\ G. Erdtmann and W. Soyka, "The Gamma-Ray Lines of Radionuclides Ordered by Atomic and Mass

Humber," J. Radioanal. Chem. 27, 137-286 (1975).

,3R. Gunnick and J. E. Evans, "In-Line Measurement of Total and Isotopic Plutonium Concentra- tions by Gamma-Ray Spectrometry," UCRL-52220, Lawrence Livermore Laboratory (1977).

205
FIGURE i| Pu AND Am DETECTION LIMITS VERSUS DEPTH IN A CONCRETE ABSORBER

o

Am

241 17J

-

L Am (Np Lo) - 59.5

0.1

D_Q

O

OCJ

10

100

1000

238 n 239 D Pu Pu 238D

43.5 239 n 51.6 Pu

DETECTOR: 19 cm PLANAR Ge

LI VET I ME: lOmin 'SOURCE - DETECTOR DISTANCE: AREA COUNTED: 66 cm2 CONCRETE DENSITY: 2.5g/cm3

AVG. AT. NO . 11 PEAKS OBSERVED

CALC'DFR. REL INTENSITY AND OBSERVED BACKGROUND

' ' liml i i i 1 1 1 1 1

1

I I I III i i i 1 1 1 iii iilinil i i ilnnt

10 10 10 1 10 10 10' 10 10'

3o DETECTION LIMITS, nCi/cm'

ACKNOWLEDGMENTS

We appreciate the assistance of H. L. Nielson in making the in-situ measurements. Thi work was performed under U. S. Department of Energy Contract EY-76-C-06-1830

.

206
Experiences With a Counter for Plutonium in Crates

by

R. A. HARLAN

Rockwell International, P.O. Box 464, Golden, Colorado 80401

ABSTRACT

Some contaminated waste generated at the Rocky Flats Plant has

been put into crates for shipment to storage and disposal. Previous

estimates of plutonium contamination in these wastes has been con- ducted with hand-held survey instruments. That practice was time

consuming and expensive in terms of manpower.

A counter to assay plutonium in 1.22 by 1.22 by 2.13 mm crates

was constructed. A previous report 3 described the counter and preliminary experiments which indicated the counter could detect 0.7 grams

of 239Pu in 6 hours or 10 grams in 20 minutes at a 0.95 confidence

level. These sensitivities assumed the plutonium would be reasonably

well distributed throughout a crate as opposed to concentrated in

the center or in a corner

.

The counter was made with 51 mm thick polyethylene walls. A

standard crate was rolled inside and measurements of instrument

response made with Pu02 sources in various locations. Ten detectors,

installed outside the walls, were lithium-loaded zinc sulfide screens

(NE-422 ) 305 mm diameter. These detectors responded to thermal

neutrons. Their scintillations were passed through conical light

pipes to 127 mm diameter photomultiplier tubes. Signals from the

tubes were processed through conventional amplifying electronics.

Discriminators were used to practically eliminate any response to gamma rays. The standards crate for the counter was used to simulate combustibles waste crates which have had a mean weight

and standard deviation of 658 ± 200 kg, respectively.

Since the previous report 3

, experiments have been conducted

to better measure the counter responses with plutonium concentrated

in unfavorable locations and responses with plutonium in noncombustibles matrices. These laboratory-type experimental results will be

described. Operating history of the counter under routine conditions

will be discussed.

Having this counter as a final assay device has the potential

of reducing assay efforts in certain plant operations. Such benefits

will also be discussed.

KEYWORDS: Analyses; counter, neutron; plutonium; waste

Sponsor: J. D. McCarthy

Work supported by U.S. Department of Energy under Contract

No. EY-76-C-04-3533

! R. A. Harlan, Journal of the Institute of Nuclear Materials

Management, 1977, in press

207
Requirements for Near-Real-Time Accounting of Strategic

Nuclear Materials in Nuclear Fuel Reprocessing

by

E. A. Hakkila, D. D. Cobb, R. J. Dietz, J. P. Shipley, and D. B. Smith

Los Alamos Scientific Laboratory, Los Alamos, New Mexico

ABSTRACT

A Purex-based nuclear fuel reprocessing plant has been studied for

possible incorporation of near-real—time accounting to supplement conventional accounting procedures. Near-real—time accounting of special

nuclear materials relies on in-line or at-line flow measurements and

plutonium assay of product and waste streams, complemented by conventional analytical chemistry for daily instrument calibrations. In-line

alpha monitors could be used for waste stream measurements of plutonium, even in the presence of high beta-gamma fluxes from fission products.

X-ray absorption edge densitometry using either K- or L-absorption

edges could be used for plutonium concentration measurements in main

product streams. Some problem areas identified in waste stream measurements include measurements of leached hulls and of centrifuge sludge.

Conventional analytical chemical methods for measuring plutonium in

weapons grade material can be modified for reprocessed plutonium.

Analytical techniques requiring special precautions will be reviewed.

Some suggested areas for improvements in process design to facilitate materials accountability in future plants will be discussed.

KEYWORDS: Nuclear safeguards; dynamic materials accounting;

in-line analysis; nuclear fuel reprocessing.

INTRODUCTION

Safeguards has become an increasingly important consideration in the public acceptanc

of nuclear energy. This is particularly true in the back end of the nuclear fuel cycle

where large amounts of fissile material are available in a relatively pure, concentrated

form. The Safeguards Systems Group at the Los Alamos Scientific Laboratory has been

tasked by ERDA and DOE to design integrated materials accounting systems for various

plants in the back end of the fuel cycle. To date, studies have been completed for plutonium handling in a LWR fuel fabrication plant, a nitrate-to-oxide conversion plant, and

a LWR fuel reprocessing plant. Each system is based on a specific facility — for the fu

reprocessing plant the AGNS plant at Barnwell, South Carolina (BNFP) was selected. This

plant is one of the most modern reprocessing plants in the world. It was designed to reprocess 1500 MT of irradiated fuel per year, producing approximately 15 tonnes of plutonium, or on the basis of 300 operating days per year, approximately 50 kg per day of

plutonium.

Design concepts and evaluation methods were developed for advanced nuclear materials

management systems to safeguard in-process materials in current and future fuel reprocess

ing facilities. The concepts are based on a thorough evaluation of the Barnwell plant an represent minor extrapolations of existing fuel-reprocessing technology, conventional

analysis, state-of-the-art nondestructive assay (NDA) , and data-processing and analysis

systems. Concepts applicable to domestic and foreign commercial or government-owned

facilities to be built in the next decade were evaluated with simulated production data.

208

:!
THE PUREX RECOVERY PROCESS

All present and currently proposed aqueous separations facilities including the BNFP

e based on the Purex solvent-extraction process, developed by the US Atomic Energy ComLssion (AEC) in the late 1940s to satisfy military ieeds for weapons-grade plutonium. In

ne Purex process both uranium and plutonium are recovered, and the radioactive waste

Diumes are reduced by minimal use of nonvolatile salting agents and reactants.

Purex and its subsequent variants are based on dissolution of the irradiated fuel in

Ltric acid, oxidation of the contained plutonium and uranium, and coextraction of the

•cidized species into an organic phase consisting of a hydrocarbon diluent containing trijjtyl phosphate (TBP) , which forms extractable complexes with the oxidized species. The

rganic phase is scrubbed to remove most of the nonextractable fission products and transranics from the coextracted fissile materials; then it is contacted with an aqueous phase

nat selectively reduces the plutonium, stripping it from the organic phase. The partiLoned phases, the organic uranium and the aqueous plutonium, are then individually subected to additional extraction, ion exchange, and other purification steps that eventually

roduce two pure product streams containing uranyl nitrate and plutonium nitrate. The

drubbing and stripping solutions from the purification and "polishing" steps are recycled

long with the spent solvent streams to give a total recovery of uranium and plutonium

roduct that can be approximately 99.9% with a net fission-product decontamination factor

E more than one million.

Solvent damage from radiation exposure in the first extraction (co-decontamination)

tage has always been an important problem in the operation of a Purex process. This

olvent degradation poses problems for analytical chemistry as well as for the process

ngineer because the dibutyl and monobutyl phosphate plutonium complexes do not behave

a extraction or chemical analysis schemes as does the normal TBP complex. After most of

|he fission products have been removed, contact time in subsequent extraction, partition,

rid purification steps is not nearly so important, and less vigorous means of attaining

quilibrium are used in the interest of reduced mechanical complexity.

Other areas receiving attention in modified Purex processes have been the selection of

uitable reductants for the partitioning step and the postpartition purification and

olishing operations. The desire for reductants having minimal effect on product purity

nd waste volumes has resulted in the use of organic reductants that decompose to volatile

roducts, of uranium-IV reductants that add no new material to the product, and, ultiately, of electrolytic reduction, which adds nothing to the process stream, as in the

roprietary AGNS Electropulse System.

BNFP is pure state-of-the-art Purex: chop-leach head end with continuous dissolution,

entrifugal extraction, and electrolytic partitioning. There are two uranium-extraction

ycles plus a silica-gel polishing column and two plutonium-extraction cycles. First-stage

xtraction is performed with the Robatel centrifugal contactor; subsequent co-decontaminaion, partition, and purification contactors are pulsed columns, including a proprietary

lectropulse partitioning column. In this system, partition is effected by electrolytic

eduction of uranium-VI to uranium-IV, which subsequently reduces plutonium-IV to plu-

,onium-III. Considerable uranium is stripped along with the plutonium, necessitating an dditional plutonium-purif ication step and significant back cycle. The plant uses remote

aintenance at the head end and through the first cycle, and anticipates no maintenance in

ther areas of the plant except those in which the products have been decontaminated to he level where direct maintenance is feasible.

Because details of the AGNS design and materials flows are available and because the

lant itself is in an advanced state of completion, it was selected for our safeguards

odeling and design purposes.

CONVENTIONAL MATERIALS ACCOUNTING AT BNFP

o The conventional materials accounting system at BNFP is based on daily analysis of the

ccountability tank, waste streams, and plutonium nitrate product tanks. The measurement

oints and precisions of analytical methods are summarized in Table I. NRC regulations

209
TABLE I

CONVENTIONAL MATERIALS ACCOUNTABILITY AT BNFP

Tank Frequency Method

Measurement

Accountability 3/day Mass spectrometry 0.2 to 0.5%

Leached hulls 1/batch 144Pr ? HAW surge 6/day Mass spectrometry 5 to 3%

Pu nitrate each batch Coulometry, 0.3 to 0.1%

amperoraetry

specify 1% error in measuring plant throughput. Throughput is verified semiannually by

f lushout-cleanout and measurement of in-process holdup. Based upon a plutonium throughput

of 50 kg per day, the allowable material-unaccounted-for between cleanouts is 75 kg.

Thus, one is faced with the problem that a divertor could hide sufficient plutonium fc several weapons in the measurement uncertainty, and it could take up to six months before

one could detect missing material. Clearly, this is an uncomfortable situation for the

nuclear materials manager.

DYNAMIC MATERIALS ACCOUNTING

The basic philosophy behind dynamic materials accounting is to provide sufficient

instrumentation that one can obtain near-real-time measurements of material flow rates

and concentrations in all streams leading into the process area. Furthermore, the process

could be divided into as many materials accounting areas as one can successfully instrument. Thus, timely material balances can be drawn about relatively small amounts of SNM. For a plant such as BNFP, the most desirable areas to instrument would be those containing

the largest amounts of plutonium in a form most attractive to the divertor. The plutonium

at the head end of the process is not attractive because it contains lethal concentrations

of fission products and is diluted approximately 100-fold with uranium. However, after

the IB column, the bulk of the fission products have been removed and the uranium/plutoniun

ratio has been reduced to 2/1. From this point the plutonium becomes increasingly attractive as it proceeds through the process to the plutonium nitrate storage tanks. Hence,

this area was selected for design of a dynamic materials accounting system.

System Design

A schematic of the BNFP is shown in Figures 1 and 2. To isolate the plutonium purification process (PPP) (Figure 2) as a unit process requires flow and concentration measurements at the IBP tank (input) and 3P concentrator (output). In addition, acid recycles

(2AW, 3AW, 3PD) and organic recycle (2BW, 3BW) must be monitored for flow and concentration. The nominal flow rates and plutonium concentrations for these streams are summarizec

in Table II.

TABLE II

CONCENTRATIONS AND FLOW RATES IN THE PPP

Plutonium

Stream Flow (L/h) Concentration (g/L)

IBP 400 5 3PCP 8 250

2AW 500 trace

3AW 215 0.1

j 3PD 32 trace

2BW 150 trace

3BW 105 trace

210
Flow Measurement

The precisions for flow measurement in the input and output streams should be 1% (la) etter. For the IBP stream this can be attained with a calibrated orifice meter in the

i surge tank or with in-line ultrasonic or magnetic flowmeters. The flow measurement

isions for the recycle streams are less stringent; possibly air lifts could be calited to the order of 5-10%.

Concentration

The plutonium concentrations of input and output product streams of the PPP can be

iured using absorption-edge densitometry. The Pu concentrations in the IBP stream can

aeasured at the L^.-j-j- edge using either x-ray tube^ or bremsstrahlung sources.

1^ For

/.onium concentrations of approximately 5 g/L a precision of 1-2% can be obtained.-^ The

stream also could be measured at the K edge using longer cells. For highly radioactive

itions a curved crystal spectrometer may be used as an energy filter for an energy dis-

. ;ive detector.-' Additional research and development is required to evaluate the effect

"'ission products on the method and to measure the precision and accuracy under plant

litions.

The plutonium isotopic composition and concentration at concentrations representative

:he 3PCP stream for reprocessing samples can be measured using a radioisotope source at K-absorption edge. Using a ^Se-^Co source as suggested by Canada, ^ Hofstetter, et obtained a precision (la) of 0.2 to 0.5% for plutonium concentrations between 150-500

The plutonium isotopic composition of these samples was similar to that expected for

st-cycle LWR fuel.

The recycle stream concentrations generally are expected to be less than 0.1 g/L, hence

:er measurement precision can be tolerated than for the product streams. In-line alpha

..tors have been installed in these streams for process control to assure that columns

: operating properly. These alpha monitors are being evaluated at the BNFP for the

;ititative measurement of plutonium in flowing streams. They have been shown to have a

:ar response to plutonium concentration in the range between 10^ to 10^ dpm/mL with an

-la/beta discrimination factor of 10^.^ A relative precision of 5-10% (la) for plutonium

-:entrations should be obtainable.

One area that presents a problem in measuring plutonium in the plutonium purification

i is the determination of in-process holdup. Capacities of the tanks and columns in

3 area are shown in Table III.

j Approximately 22 kg of the 41-kg holdup is in the IBP tank and the concentrator, and

be estimated from the in-line concentration instruments and tank volumes. However, the

-int in the columns cannot be measured accurately, and can contribute significantly to

1 uncertainty in plutonium content of the PPP.

An alternative to actually measuring plutonium concentration is the estimation of conizations from a knowledge of column operation.^ The feasibility of this approach was

:ed by mathematical simulation of column operation for the 2A column. A schematic of

2A column, with flow rates into and out of the column, is shown in Figure 3. The

:onium holdup as a function of 2AF, 2AS, and 2AX flow rates was modeled using the SEPHIS

and results are summarized in Figure 4.^ The largest variations in holdup occur

llow 2AX (organic) flow rates. However, the data indicate that from a knowledge of the

at stream flow rates the plutonium holdup in the column can be estimated. The accuracy

-he estimate has to be verified with actual column operation.

DISCUSSION AND RESULTS

The operation of the plutonium purification process was mathematically modeled using

te Carlo techniques. ^ The measurement points and associated errors for each point are

iiarized in Table IV. Four different strategies were used to evaluate the diversion

sitivity, and are summarized in Table V. The measurement errors for the four cases are

211
TABLE III

IN-PROCESS HOLDUP IN TANKS AND VESSELS OF THE PPP

Identification

Volume

(L)

3 lutonium

Concentration

(g/L)

Plutonium

Holdup

(kg)

IBP tank

2A column

2B column

3A column

3B column

3PS wash column

3P concentrator

1500

700

500

600

440

20

60

4.942

b

b

b

b 58.70

250.

7.4

4.6

2.8

5.4

4.8

1.2

15.

See Figure 2.

A model of the concentration profiles and the holdup in the pulse columns

is described in Reference 9.

These values are not flowsheet values of any existing reprocessing facility

but represent typical values within reasonable ranges of a workable flowsheet .

TABLE IV

MEASUREMENT ERRORS FOR DYNAMIC MATERIALS ACCOUNTABILITY IN THE PPP

Precision Calibration Error

(lo), % (la), %

IBP Flow 1 0.5

Concentration 1 0.3

2 AW Flow 5 1 2BW

3AW Concentration 10 2 3BW

3PCP Flow 1 0.5

Concentration 1 0.3

TABLE V

MEASUREMENT STRATEGIES

In-Process Inventory

Balance Recalibrate Measurement Precision

Case Period Flow (la), %

1 8 h — 10

2 8 h 24 h 10

3 8 h 24 h 5 4 1 h 24 h 5

Measurement of concentration and flow every 0.25 h. 212
rized in Table VI, and show that in each case measurement error is dominated by uncerInties in in-process inventory. The diversion sensitivity was determined using decision

lysis theory described in a preceding paper ("Decision Analysis for Dynamic Accounting

Nuclear Material" by James P. Shipley), and is shown for cases 1 and 4 in Table VII. rovement in diversion sensitivity is obtained by increasing measurement frequency. Note

t for case 4 (1-h measurement frequency) a diversion sensitivity of 4.2 kg at the end of

week is obtained. This can be compared to the conventional sensitivity of 75 kg per

-month inventory period.

TABLE VI

MEASUREMENT ERRORS IN FOUR DYNAMIC-ACCOUNTABILITY CASES

Variance (kg Pu) (standard deviation, kg Pu)

! transfers

ne-week average

process inventory

ne-week average

erial balance

i'ne-week average

urn nd of day

,nd of week

Case 1 Case 2 Case 3 Case 4

0.022 (0.15)

0.98 (0.99)

1.99 (1.41)

2.13 (1.46)

8.51 (2.92)

0.022 (0.15)

0.98 (0.99)

1.99 (1.41)

2.13 (1.46)

3.29 (1.81)

0.022 (0.15)

0.36 (0.60)

0.74 (0.86)

0.89 (0.95)

1.98 (1.41)

0.00073 (0.027)

0.36 (0.60)

0.74

0.89

1.98

(0.86)

(0.95)

(1.41)

TABLE VII

DIVERSION SENSITIVITY FOR THE PLUTONIUM PURIFICATION PROCESS

(8 h)

(1 h)

Average Diversion

per Balance (kg Pu)

4.2

0.30

0.15

2.6

0.075

0.025

Detection

Time (h)

168 (1 week)

672 (4 week)

1 24 168 (1 week)

Total at Time

of Detection (kg Pu)

4.2

6.3

12.6

2.6

1.8

4.2

is determined consistently with ERDAM Appendix 7401-C, "Nuclear Materials Management and

Safeguards System Handbook."

The safeguards materials accountability improvements described in this report are based

a measurement overlay for an existing reprocessing plant. Hopefully, the system could

improved if it could be incorporated into the plant design at an early stage. The folding were identified as areas of safeguards concern for future facilities of this type.

Location of Centrifuge

One source of sampling error at the input accountability tank results from the suspended

rticulates in the dissolver solution. Solids could account for as much as 0.3% of the

:ountability-tank volume and 0.8 kW of heat per tonne of dissolved fuel. Consideration

mid be given to installing the centrifuge between the dissolver and the accountability

ik, as has been done for the Japanese plant at Tokai and the proposed EXXON Nuclear Comiy plant.

213
Accountability Tank

Further consideration should be given to using load cells to measure the volume of

input accountability tank. The design must, as much as possible, isolate the tank froir associated piping. If practicable, mass measurements using load cells would be particu

ly advantageous for processing fuels with higher burnup or shorter cooling times, which

would have higher intrinsic heat-generating capacities.

Flowmeters

Flowmeters having provision for periodic recalibration should be installed in speciUj

crucial process streams; a measurement accuracy of 1% or better is required. A 5-10% ir M

urement accuracy is acceptable for flowmeters in waste streams.

Concentration Sensors

In-line or at-line detectors should be incorporated to measure plutonium concentrat.i ri in major process streams with a measurement accuracy of 1% or better, and in waste stre with an accuracy of 5-20%.

Instrument Accessibility

All in-line or at-line instruments, including flowmeters and concentration sensors, should be installed in a manner that permits ready accessibility for recalibration and

maintenance by plant personnel or inspection by the national or international safeguard

staff. Sensors should be directly interfaced to the safeguards computer system for dynjii .. materials accounting.

HA Contactor

An improved decontamination factor may be attainable at the front end of the separa

tions process by increasing the number of stages in the HA contactor, or by providing a second decontamination cycle before partition, as has been done effectively in some oth

facilities. The reduced radioactivity in the uranium-plutonium product stream might pel- mit inclusion of an additional accountability point before the plutonium-process area a the HS column. This would be highly desirable both for process control and for safegua.Jls

in the event that a co-processing mode of plant operation is selected.

3P Concentrator

Concentration of the final PuClS^)^ from 60 to 250 g/L is primarily for convenience

storing and shipping. Under current NRC regulations co-location of reprocessing and

nitrate-to-oxide conversion facilities will be required. If the concentration of the fifes plutonium-product solution from the reprocessing plant is maintained below 50 g/L, thisl

solution can be used directly as feed for the conversion plant, and the 3P concentrator ii associated heater and feed tanks can be eliminated. From a safeguards viewpoint, this

would increase the amount of solution required to divert 1 kg of plutonium by a factor , 4 to 8, and would decrease holdup by the volume of the concentrator. In order to provi

a 3- to 6-month product-storage capacity, the number of nitrate-storage tanks would hav

to be increased proportionally unless the plutonium-product output were directly couple

to the input of a contiguous conversion plant.

Centrifugal Contactor

In general, in-process inventory can be reduced by a factor of 20 by using centrifu

contactors in place of pulsed columns in the plutonium-purif ication area.^ Using a French-designed eight-stage contactor, the pulsed columns in the plutonium-purif ication

214
j.rea could be replaced as follows:

2A column - 9 contactors

2B column - 3 contactors

3A column - 4 contactors

3B column - 2 contactors.

in addition to low holdup, centrifugal contactors provide added advantages of rapid draintown with negligible drain-down volume, and rapid startup after shutdown. Disadvantages

nclude relatively low flow rates of 100 L/h in both aqueous and organic phases, greater

Lccessibility of material to possible diversion, and more stringent design requirements for

ptimizing organic/aqueous flow ratios.

CONCLUSIONS

The primary conclusion from this study is that current technology can provide improved

rocedures for safeguarding strategic quantities of SNM in a nuclear-fuel reprocessing

acility at a reasonable cost and with minimal disruption of production processes. The

ystem design must be facility-specific for each reprocessing plant, taking into consideraion such features as plant throughputs, side streams, materials control philosophy, and

quipment maintenance features. The formation of a dynamic materials balance area around

he plutonium-process area adds another level of safeguards protection of plutonium in its

ost concentrated and pure form, the form most attractive to a potential divertor.

The study has also identified generic features and processes in the reference facility

hat contribute most importantly to measurement uncertainties and that could be improved in

uture facility designs if they were to be optimized for their safeguardability

.

REFERENCES

"Barnwell Nuclear Fuels Plant Separations Facility—Final Safety Analysis Report,"

Docket 50-332, Allied-General Nuclear Services, October 10, 1973.

E. A. Hakkila, D. D. Cobb, H. A. Dayem, R. J. Dietz, E. A. Kern, E. P. Schelonka, J. P. Shipley, D. B. Smith, R. H. Augustson, and J. W. Barnes, "Coordinated Safeguards for

Materials Management in a Fuel Reprocessing Plant," Los Alamos Scientific Laboratory

report LA-6881 (September 1977).

E. A. Hakkila, R. G. Hurley, G. R. Waterbury, "Three-Wavelength X-Ray Absorption Edge

Method for Determination of Plutonium in Nitrate Media," Anal. Chem. _38, 425-427

(1966).

T. R. Canada, S. T. Hsue, D. G. Langner, E. R. Martin, J. L. Parker, T. D. Reilly, and

J. W. Tape, "Applications of the Absorption-Edge Densitometry NDA Technique to Solutions and Solids," Nucl. Mater. Manage. VI, No. Ill, 702-710 (1977).

J. W. Tape, T. R. Canada, and E. A. Hakkila, "Application of Dispersive X-Ray Filtering

Techniques to Absorption-Edge Densitometry," in Nuclear Safeguards Research and Development Program Status Report January-April 1977 , Los Alamos Scientific Laboratory

report LA-6849-PR (August 1977), pp. 10-11.

T. R. Canada, J. L. Parker, and T. 0. Reilly, "Total Plutonium and Uranium Determination by Gamma-Ray Densitometry," ANS Trans. 22^, 140-142 (1975).

K. J. Hofstetter, G. A. Huff, R. Gunnink, J. E. Evans, and A. L. Prindle, "On-Line

Measurement of Total and Isotopic Plutonium Concentrations by Gamma-Ray Spectrometry,"

Twenty-First Conference on Analytical Chemistry in Energy Technology, Gatlinburg,

Tennessee, October 4-6, 1977.

215
;

8. K. J. Hofstetter, G. M. Tucker, J. H. Gray, and G. A. Huff, "Applications of On-Line

Alpha Monitors to Process Streams at the Barnwell Nuclear Fuel Plant," 1978 Spring

Meeting, American Chemical Society, Anaheim, California, March 13-17, 1978. To be

published in American Chemical Society Symposium Proceedings Series.

9. D. D. Cobb, "Appendix G: Dynamic Model of the Plutonium Purification Process," in

Coordinated Safeguards for Materials Management in a Fuel Reprocessing Plant , Vol. II Los Alamos Scientific Laboratory report LA-6881 (September 1977).

10. G. W. Reddick, "Solvent Extraction in HTGR Reprocessing Interim Development Report,"

General Atomic report GA-A13835 (1976).

11. B. M. Legler, "Sensitivity Analysis for Plutonium Holdup in the 2A Column," personal

communication, May 4, 1976.

12. A. F. Cermak, Allied-General Nuclear Services, personal communication, August 15, 197

216
FEED SURGE TANK

Z

ACCOUNTABILITY TANK

*

H A F FEED TANK

Figure 1. Dissolver-separations process block diagram.

217
Figure 2. Plutonium purification process block diagram.

218
2 AS

60 L/h

2 AF

430 L/h

_ 2AX__

~I45 L/h"

2AP

2 AW

Figure 3. Schematic of 2A column.

219

tification of 235U Enrichment in SNM Receipts at the DOE Feed Materials Production Center

by

H . W. HUMPHREY

National Lead Company of Ohio, Cincinnati, Ohio

ABSTRACT

A diversity of uranium-bearing, nonirradiated materials with 23

5

U assays

from fully depleted to a maximum of 10 wt %

23

5

U, is routinely returned

to this DOE site for reprocessing. It is necessary to confirm the 23

5

U isotopic content of Special Nuclear Material (SNM) receipts both for

accountability and nuclear criticality safety purposes.

Methods and instrumentation have been developed for an isotopic verification facility by which 23

5

U isotopic content is rapidly and nondestructively determined. The 185 keV gamma photon emitted by 23

5

U is measured

using a dual-channel gamma spectrometer which incorporates automatic

background subtraction and an optimized collimator /Nal (Tl) detector

assembly. Measured activities are then related to 23

5

U isotopic content

by means of appropriate standards that have been destructively analyzed

in the laboratory.

Use of the described facility for the past seven years has resulted in

improved nuclear safety procedures, better use of storage space and more

stringent nuclear materials accountability.

KEYWORDS: Gamma spectrometry; 23

5

U enrichment; nondestructive analysis > sodium iodide detection; isotopic verification

INTRODUCTION

2 National Lead Company of Ohio is the contract operator of the Department of Energy (DOE)

ad Materials Production Center (FMPC) . The FMPC is located near the village of Fernald,

sroximately 20 miles northwest of Cincinnati, Ohio. Nonirradiated uranium materials,

th 235U enrichments of less than 10 wt %, are routinely returned to the FMPC for recovery

uranium, which is subsequently returned to the nuclear fuel cycle.

the past difficulties have been encountered with scrap receipts in the areas of both

2cial Nuclear Materials (SNM) accountability and criticality control. Some of these

cficulties were previously reported. 1 These problems are the results of several things:

strumentation may not be available to perform 235U isotopic assays and calculated values

2 therefore used; in other cases 235 U assays are not obtained because enrichment values

2 thought to be reasonably documented; another, and perhaps the most common reason for

isotopic discrepancies is human errors involved in handling both the material and its

';ompanying paperwork.

• L. Dunaway. "Enrichment Errors in Interplant Shipments". USAEC Report NLCO-1083

Tune 29, 1971).

221
Since it was felt that these problems could continue in the future, and their consequenc

could have serious nuclear criticality potential in our processes, a program was initiat

to check the 23

5

U isotopic content of virtually all incoming SNM receipts. Rigid materi

handling and operational procedures have been adopted to preclude the accidental (or sui

titious) loss of SNM's or the acceptance of highly enriched materials by mistake.

EXPERIMENTAL

It became apparent several years ago that a rapid, nondestructive means of verifying th< 23

5

U isotopic content of uranium receipts was needed. Requirements for the method were

it should be inexpensive and relatively simple to operate,

operate the instrumentation on a day-to-day basis.

Nontechnical personnel were

Theory of Operation

A technique based on the measurement of the 185 keV gamma photon, emitted by 23

5

U (54% < the time) 2

, was chosen as the most economically and operationally acceptable method. T approach finally adopted was a refinement of a previously described method. 3 It can be

easily demonstrated that the 235U isotopic content is directly proportional to the meas> 185 keV activity, for uranium in which the measured thickness greatly exceeds the 185 k

e-f old attenuation distance (infinitely thick for our purposes)

.

Given the geometric configuration of Figure 1, we readily conclude that the incremental

count rate "d(cr)" arriving at the Nal(Tl) detector from uranium metal can be described

d(cr) = eripA exp[-(L-x)/X]dx (1)

where,

cr =

e = L =

X =

r =

i =

p = A =

V =

count rate (c/s)

detector efficiency for 185 keV gamma rays

slab thickness (cm)

e-fold attenuation distance (cm) for 185 keV

gamma rays in uranium (mean-free-path)

specific 185 keV gamma ray activity

(4.255 x 10u gamma rays/s - g 235U)

23

5

U weight fraction

uranium density (g/cm3

) area viewed by the detector (cm2

) uranium linear attenuation coefficient (cm-1

)

Integrating over the slab thickness L yields

cr = enP AX(l-e"L/X

) (2)

assuming L >>X, attentuation with regard to spatial distances can be disregarded and

cr = eripAX (3)

or, since X = 1/up, and rearranging terms we have:

I = Aef

(cr) (4)

"Table of Isotopes", 6th ed. 2C. M. Lederer, J. M. Hollander, and I. Perlman.

New York: Wiley (1967).

3 B. L. Twitty and H. W. Humphrey. "A Self-Correcting Inspection Monitor for Determinir

the 23

5

U Level in Reactor Fuel Cores". USAEC Report NLCO-1011 (May, 1968).

4 R. B. Walton.

Los Alamos, N. Personal Communication to

M. (March 5, 1971).

H. W. Humphrey from R. B. Walton of LASL,

222
t is interesting to note: (1) that in this form (Equation 4) the 235U isotopic content does

ot depend explicitly on the density, and (2) all the terms in the first portion, of the

jlght side of the equation are constants, therefore the 23

5

U isotopic content is directly

roportional to the measured 185 keV gamma activity.

nf ortunately , most of the materials received at this site are not merely uranium metal,

ranium in practically every conceivable geometric configuration and chemical composition

as been returned for reprocessing at one time or another. Some of the materials received

3r reprocessing are listed in Table I. The listing does not, of course, include all the

Dims and matrices returned to this site for reclamation of uranium values. The physical

intensions of receipts have varied from several mils (uranium wire) to large railroad tank

ars. Uranium concentrations ranging from several hundred ppm (contaminated copper) to are metal (100% U) have been encountered.

ince the matrices to be measured are quite varied, a means of correcting for geometry,

Dntainer (cladding), and composition is necessary. Variations in both size and shape can

"b minimized using a series of graded lead collimators. The largest size collimator is

ormally used, consistent with both the surface area of the specimen and the ability to sproducibly place the specimen on the detector. The greater the area viewed, the greater

ill be the measured count rate (hence the more precise will be the reading)

.

1 mathematical model has been formulated to compensate for heterogeneous regions (voids)

Drmed by nonuranium materials localized in the detector viewing area. Since the procedure

3 based on the assumption of a point source of radiation, working relationships should be

itermined empirically using actual broad-beam conditions.

le inverse square law states that radiation intensity is inversely proportional to the

}uare of the source-to-detector distance (D) . The ratio of activity measurements (A)

and "b" are then given by

A(a)/A(b) = D(b)

D(a) (5)

I calculations were made at one inch increments and the results plotted, the lower (no void)

arve shown in Figure 2 would result. Moving the source 1/2 inch upward to simulate the

Lsplacement due to a 1/2 inch void, and so on for the other void thicknesses, would yield

ie other indicated relationships. To use the algorithm, one determines the activity ratio

)r an unknown material at two different source-to-detector (S-T-D) distances. Referring

) Figure 2 should provide the approximate void size (if any) ; the appropriate void correclon factor can then be determined using the upper family of curves.

: should be noted that the affect of voids within a container can be minimized by judicious

ljustment of the S-T-D distance. For instance, if we measure the activity of a point

>urce located 1 inch from a detector and then move the source 1-1/2 inches away from the

:tector and take a second measurement (simulating a 1/2 inch void) the activity ratio would

Ai.s / A1 = 0.444 (6)

ipeating the measurements at 12 inches and 12-1/2 inches we find

A12.5/A12 = 0.922 (7)

, less than 8% error is introduced into the measurement at a S-T-D distance of 12 inches,

a 1/2 inch void is present.

1 practice, the selection of the S-T-D distance is usually a compromise. This is seen more

.early using the above example. Increasing the S-T-D distance from 1 to 12 inches reduces

ie activity reaching the detector by more than 99%. This of course means that measurement

:::jnes must be increased to insure adequate counting statistics. The minimum S-T-D distance

defined in practice by the thickness of the collimator employed.

223
In addition to the 185 keV photon emitted by 235U, most uranium compounds also emit radiations of 765 keV and 1001 keV (exceptions would be freshly purified materials). These tw< gamma rays are emitted during the decay of 23AmPa (UX 2 ) > the granddaughter of 238U, and

account for most of the sample background via Compton interactions. The 235U photopeak

rests upon this Compton continuum from higher energy photopeaks. The collimator thicknes

is therefore selected to maximize the difference between 23

5

U and 234mPa transmissions

(Figure 3). Lead was chosen for the collimator material due to its high density, availab

and malleability. Sodium iodide (thallium activated) was the detector of choice due to i high gamma ray efficiency, relative availability, simplicity, and low cost. The detector

thickness was optimized by differentiating the expression for the difference between the

two transmission curves (Figure 4) and setting the first derivative equal to zero to find

the maximum. An optimum thickness of 0.53 inch was thus calculated.

Although spatial effects play a very important part in nondestructive gamma spectrometry,

equally important (and often overiding) effect is the sample composition itself.

It was shown earlier that for uranium metal, the 185 keV count rate was independent of th density of the matrix emitting the radiation. In the case of uranium compounds and resid

however, this is not the case. Equation (4) must be modified to include the effect of th nonuranium components of the matrix. Quantitatively this becomes 5

where,

I =

F =

AeY

1 +

(F)-(cr)

m m

MP

(8)

(9)

ymmand p m are the mass absorption 1 coefficient and densityj of the nonuranium matrix mater: respectively, and y and p are as previously defined.

The mass absorption coefficient is a slowly varying function of atomic number6 up to aboui

Z = 30; beyond that it is seen to increase rather rapidly (Figure 5) . Although the chem: composition of receipts is not always known, quite often it is not required. From Figuri we see that the ratio of um/y ;0.08 for Z = 30. If pm/p < 1.0 (50% U or more) the affec

upon the analysis would be an error of less than 10%. In practice this means that the 2 isotopic content of uranium compounds and rich residues (> 50% U) can be determined by

comparison with a single standard without making a correction for matrix absorption. If the matrix and uranium concentration are known, "F" can be evaluated; if a standard and

unknown are of the same material, "F" may be neglected. The technique is not applicable

materials containing high Z elements in unknown concentrations or to mixtures containing

less than an "infinite" thickness of material, unless the values of these parameters are

known.

An "infinite" thickness of material "T" , for purposes of this assay, is defined as that

quantity (thickness) necessary to give 95% of saturation. Quantitatively this becomes

10 In 0.05

I = mm

- y (10)

Infinite thicknesses for several of the more common uranium compounds are listed in Tabl

5J. L. Parker. Personal Communication to H. W. Humphrey from J. L. Parker of LASL,

Los Alamos, N. M. (December 31, 1970).

6J. H. Hubbel. "Photon Cross Sections, Attenuation Coefficients, and Energy Absorption

Coefficients from 100 keV to 100 GeV". NSRDS-NBS-29 . (August, 1969).

224
jation (5) must also be modified to reflect the effect of container attenuation. The net

:ect of container attenuation is to lower the measured count rate. If the container (or

adding) thickness is known (or can be measured) the container activity transmission is

Lculated using the relationship

C = Coe^ct (11)

are,

Co =

C = ^c "

t =

The unattenuated count rate (c/s)

The attenuated count rate (c/s)

Linear absorption coefficient of the

container material (cm-1

) Container thickness (cm)

:orporating this into equation (8) we have

y

/ x y c

t (cr) e

L

ymPi l +

yp

(12)

effect of container thickness upon the measurement of 23

5

U enrichment is vividly

.ustrated in Figure 6. Metal thicknesses of greater than 0.025 inch produce results that

low by 10% or more.

above analytical expression (Equation 12) serves as the basis for the measurement of

;nown enrichments. In practice, however, there are other consid erations that must be

en into account.

Instrumentation

: the gamma activity of an aged uranium sample is measured with a Nal(Tl) detector, a

iictrum similar to Figure 7 is obtained. As seen in the figure, the 235U peak rests upon

sizeable Compton continuum. At equilibrium, the contribution of 23Z* mPa to the 235U peak

a is fairly constant. If the uranium is freshly purified, the dotted spectrum of Figure

rould result upon gamma activity measurement. Following purification, the 23Z,mPa activity

i a uranium matrix will continue to grow for 5 or 6 months until equilibrium is once again

ched (Figure 8). Since the 23£* mPa contribution to the 235U peak area is also dependent

enrichment, geometrical configuration and composition, a Compton background subtraction

t be made. Several different approaches have been taken by various analysts to determine

s background contribution, for as many different reasons. The method chosen for our use

to perform a Compton background measurement in the area immediately adjacent to the 185

1 photopeak area, on the high energy side.

hough a single channel analyzer (SCA) system would be the simplest and least expensive,

s type of system requires more work from the analyst and provides additional opportunis for error. A dual channel analyzer (DCA) system (Figure 9) was ultimately chosen since

was still relatively inexpensive and did not suffer from the aforementioned deficiencies

the SCA system.

J'

i of the electronic components of the system were manufactured by Harshaw Chemical Co. and

I of the NIM type (except of course for the integral probe assembly) . The NJ-20 Up-Down

Titer automatically subtracts the background from the gross count yielding the net count

its output. The total cost of the instrumentation was approximately $2,950 when

chased in 1970.

225
IVF Facility . The above DCA system is permanently housed in the southwest corner of

warehouse near the receiving dock which is dedicated to temporary storage of SNM receipts.

The nuclear instrumentation and associated materials handling equipment are referred to by

personnel at the FMPC as the Isotopic Verification Facility (IVF).

Figure 10 illustrates the layout of the IVF. A conveyor is located behind the monitoring

table which permits large objects (e.g. drums, ingots, etc.) to be moved from the outside j and positioned above the Nal(Tl) detector with the aid of the overhead hoist. The table c the right is adjustable to accommodate long, or irregularly shaped objects. A calibrated

vernier under the table permits reproducible adjustments to the height of the radiation

detector. A large overhead door permits easy access to the IVF by fork lift trucks to br:. in bulky materials or larger quantities of receipts. Provision has been made to change

collimators on the Nal(Tl) detector to permit analysis of the various geometrical configuiJ

tions received. The temperature of the IVF is controlled year-round to approximately + 5

( to insure instrumental stability.

Support Instrumentation . The weight and piece count of each shipment is verified jui

outside of the IVF in Building #71, prior to 235U enrichment verification. For nuclear

safety purposes, materials are stored in the same configuration in which they were shipper

until the 23

5

U content can be verified.

Should it be necessary or expedient to monitor 235U enrichments in the field, a recently

acquired portable Eberline dual channel gamma spectrometry system (SAM-II) is employed.

This instrument employs a two inch diameter by one-half inch thick Nal(Tl) scintillation

crystal embedded in a lead collimator /shield probe assembly (RD-19) . The Nal(Tl) crystal

is doped with 241 Am to provide instrumental stabilization in the presence of electronic o ambient environmental changes.

If materials, or assays, are suspect for any reason, corroborating analyses are performed

at the analytical laboratory. A 20 cc Ge(Li) detector system is available to perform non destructive gamma analyses. This instrument, mainly because of its superior resolving

ability, has proven to be extremely useful over the years in diagnosing spectral difficul

In addition, where materials can be sampled, destructive gamma and mass spectrometric

analyses are often used.

Verification Procedures . All verification procedures are governed by FMPC Standard

Operating Procedure l-C-908 "Checking the 23

5

U Isotopic Identification of Materials

Received from Off-Site", dated January 12, 1973.

Both analyzers are calibrated each day prior to use. The calibration procedure requires | approximately 15 minutes per day and involves taking 20 second measurements at varying

threshold readings on each SCA with a relatively small window setting. A completed calit

tion sheet is illustrated in Table III and plotted in Figure 11. The data points marked

L, P, and L' represent the low, peak, and equivalent threshold points, respectively. Sir; the shape of the spectrum is dependent on the enrichment, material composition and geomet

the uranium calibration should be conducted using a specimen from the shipment to be veri

It will be observed in Figure 11 that there is a 0.05 difference in threshold settling be tween the two analyzers. This difference is taken into account when the background settlj

is determined for #2 SCA as seen in Figure 12. The gross count threshold "E", is simply j; "Lx"; the gross count window is

AEi = Lj - L a (13) ;

The background count threshold is (Ll-0.05) and the background window AE 2 = AEi. Having

established the calibration, the instrument can now be standardized.

226
:lardization normally takes one of several forms: (1) if all of the units within a shipare similar in physical and chemical properties and have the same stated enrichment, a element statistical standardization is performed, (2) if the units of a shipment are

2ved to be similar, but with varying enrichments, several representative units are

:ted for use as standards, (3) shipments with varying uranium concentration and enrichrequire additional information to effect an assay. Verifications can often be performed

;,iaterials in this latter category using assumed uranium compositions from shipping docuselecting the appropriate collimator, materials with the same chemical and physical

irties as well as identical 235U enrichments are standardized after randomly selecting

tits from the group. The net count rate is then determined for each unit to a coeffi-

.: of variation of ±1.5% (3a), or better.

.ength of time necessary to attain a relative counting precision of ±1.5% is a function

»th the gross 235U count "G" and the Compton background count "B" . Since the net count

.s

N = G - B (14)

1 associated with the net calculai

igating the errors involved in both measurements. Or,

°N

= ±V^T (15)

:ince for a random Poisson distribution the precision can be estimated by

aG

= ±Vg~ aB = ±Vi~ (16)

la (99%) precision associated with a net activity measurement is then

3aN

= ± 3~\/G + B (17)

ifficient of variation of 1.5% is attained when

, c _ 100 x 3 V G + B , 1Q . 1,5

(G - B) (18)

•.fining k = B/G, the gross count necessary to obtain a coefficient of variation of 1.5%

G = 40,000 ^

+

(19)

ength of time necessary to attain a 1.5% counting precision is seen to double (using

:ion 19) as the background increases from 20% to 50% of the gross count.

:g measured the net count rate of 10 randomly selected units from the shipment, the one

-which most nearly represents the average for the group is selected as the standard,

'.lly, since there are so many undefined variables, a small specimen is taken from the

.ard and destructively analyzed. All subsequent assays are then compared to the remain-

•ortion of the standard in the following manner. Ten consecutive net activity measurei are taken for the standard and the dispersion of the data determined at the 99% (3a)

.dence level. The upper and lower acceptance limits are then simply

UAL = x + 3a (20)

LAL = x - 3a

227
where "x" is the average value computed for the ten measurements. An unknown activity

measurement "Cx

" is acceptable when

LAL £ Cx

<_ UAL (21)

The 235U isotopic enrichment corresponding to count "Cx

" is calculated from the simple

tionship

w- (22) 1

where,

I = 235U isotopic enrichment (wt %) C = Integrated count rate

s = standard

x = unknown

Occasionally, when the number of units within a shipment is prohibitively large, or the history of the material suggests a uniformity of 23

5

U enrichment, a relatively poorer

precision will be found acceptable. For example, a recent shipment contained 4,123 ni

clad uranium plates enriched to 1.865 wt % 235U. The average net count rate for the

standard was 11,850 counts/15 seconds and the 3a precision was ± 0.108 wt % 235U (5.79

coefficient of variation). Had this shipment been counted to a relative precision of

a 2.5 minute count per unit would have been required. Operationally this would have

required 170 mandays, compared to the 17 mandays actually used.

The decision to either accept or reject an assay is based upon a predetermined, object

criterion. If the shipper's claimed 23

5

U enrichment agrees with the determined value

within ± 0.5 wt % 235U (99% C.L.), it is accepted. Obviously an accuracy of ±0.5 wt % would be unacceptable from an SNM accountability standpoint since large MUF's could re from such a procedure. Although the shipment is accepted if within ± 0.5 wt % 235U, t actual nondestructive assay is performed to a precision of ± 0.1 wt % 235U, or better,

this way questionable results are identified and destructive assays are performed wher

i

:

-

i

Bit

indicated

.

ill?

Figure 13 illustrates the nuclear safety implications of the ± 0.5 wt % 235U acceptanc

criterion. Curve "A" represents the safe mass 7 storage relationship; curve "B" demons

the critical mass for the indicated enrichments. Curve "C" represents the effect of s materials at a safe mass, for a stated enrichment of "I", when the enrichment could ac be (I + 0.5) wt % 235U. It will be noted from the figure that for enrichments below 1 % 235U the storage mass would exceed the critical mass, if the case of maximum error Wj assumed. This difficulty can be overcome in several different ways.

All materials containing less than 1.55 wt %

23

5

U can be stored at the safe mass for ill enrichment (approximately 360 lbs) , or less restrictive criticality parameters may be !«f as a means of control (if practical). Other means of criticality control, in additiorfafl mass, are: slab thickness, cylinder diameter, container volume, concentration, density^

mass per unit area.

A second type of standardization is employed for materials of similar chemical and phy.c<

properties but with varying 235U enrichments. For this type of shipment a quite accuisj

standardization can be performed by selecting several units that will be used as stanch

7D. L. Dunaway. "FMPC Nuclear Safety Guide". Unpublished. (Revised January 15, 1968 j

228
acket the anticipated range of 235 U values. A linear regression is then performed

the method of least squares. Raw data is fitted to an equation of the form

y = a c (23)

I,

x = wt % 235 U y = net counts/3 minutes

a 0 = y-intercept

a^ = slope

Coefficient of Determination "r 2 " is calculated to obtain an objective determination of

;oodness-of-f it of the data to the assumed regression form (r 2 = 1 is a perfect fit)

.

:ample standardization for K-25 oxides is given in Table IV. All of the computations

erformed using a programmable hand-held calculator. Analysis of the data contained in

! IV indicates that an assay could be performed to a relative precision of ± 2.1% (or

;r) in the 23

5

U enrichment range indicated.

lents of materials with both uranium and 235U variations present an especially difficult

Incinerator ashes are a good example of the type of shipment which requires special

'ment. Uranium concentrations in this type of material have been found to vary from as

s a few tenths of a percent up to almost pure U0 2 (88.15% U) . Accompanying this large

\um variation is a companion 235U enrichment variation from < 1 to 10 wt % 235U. As

earlier in equation 12, the 23

5

U enrichment measurement is dependent on the matrix

and composition; further information must therefore be available before a 235U assay

e effected when dealing with materials with such divergent uranium values.

tion that is always available is to measure the uranium concentration (either destruc-

^y, or nondestructively) 8 or perhaps perform a density measurement. 9 The net result of

' type of approach is to perform two assays for each item received.

rly novel and inexpensive semiempirical approach is taken at the FMPC to the measureof enrichment in this type of material. This approach is based on the assumption that

ally all receipts falling into this category consist of matrices with atomic numbers of

less, and the mass attenuation coefficient is therefore fairly constant.

' U3 0 8 standards were selected with known 23

5

U and uranium content which were representa1 of the full range of uranium values normally encountered (viz, 5.67%, 19.83%, and

I U).

; gamma ray attenuation is exponential in nature, a calibration equation of the follow-

'orm was assumed:

I = A c + ea + bx (24)

I = wt % 23

5

U enrichment

A = net 185 keV activity (usually per 3 minutes)

x = % uranium concentration

a,b,c = calibration coefficients

. Parker. "Nuclear Safeguards Research and Development-Status Report". USAEC Report

605-MS, p. 36. (September - December, 1970).

J. Parker, et al. "Nuclear Safeguards Research and Development-Status Report". USAEC

rt LA-4705-MS, p. 12. (January - April, 1971).

229
If L, M and H refer to the low, intermediate, and high uranium standards, then the cali

tion coefficients are found from these relationships:

In b = yL

- c (25)

where,

XH

a = ln(y

L

- c) - x

L

b

is found by iteration using the intermediate standard value of

A seed value of 1.3 x 10~ 5 is assumed for "c" and it is either incremented or decrem

units of 1 x 10-6 until y^ is attained, correct to six significant decimal places. All

j

standardization calculations are performed electronically using an HP-9810A programmablj

calculator.

Normally the assay is standardized once per day, following calibration, and all subseqvfj

measurements that day are normalized to this initial standardization using the count ra\

the intermediate standard (M)

.

Since 235U isotopic segregation has frequently been observed within containers, a measiljj is made of both the top and bottom of each container. This segregation was confirmed t! Mobile Nondestructive Assay Laboratory (MONAL) when it was deployed by LASL at the FMP(

November, 1970. 8

In order to effect a 235U enrichment measurement using Equation 24, we assume the vendc

claimed uranium value to be correct. This approach cannot be used with impunity since

stated uranium concentration is not always correct. In the absence of additional and

sophisticated instrumentation, this semiempirical approach works quite well for the ma;, of SNM receipts falling into this category. An example of a completely automated calc

is illustrated in Table V. Should differences exceed the ± 0.5 wt % 235U acceptance

criterion, messages will be printed indicating either segregation, or a discrepancy wi

vendor's claimed value.

DISCUSSION AND CONCLUSIONS

With the advent of recent terrorist activities, the need for strict safeguards and SNM

accountability have taken on new emphasis. Nondestructive SNM measurements are becomi

increasingly popular as a viable means of control throughout the entire nuclear fuel c

Although the risks of potential diversion of SNM's are not as great at the FMPC as at ties handling highly enriched, weapons grade materials, our problems are very similar. [1 natural uranium can be made critical, given the proper conditions of geometry and neut

|

reflection.

The FMPC is licensed to handle materials containing up to 10 wt % 235U. ments have been sent to us containing greater than 10 wt % 23

5

U. However , some,fl

J. L. Parker. "Nuclear Safeguards Research and Development-Status Report". USAEC Re

W

LA-4605-MS, p. 36. (September - December, 1970).

230
;jLl over 40,000 items have been nondestructively examined during the seven years since the

:eption of the IVF, with exceptionally little down time. The simplicity of the system has

ioubtedly contributed to its reliability and longevity.

recent years the IVF instrumentation has been used for purposes other than originally

;ended. Such uses as inventory verification, sorting of uranium into scrap categories

;ed on its uranium content, identification of transition points in enclosed systems, etc.

7e become increasingly more common.

; of the instrument has resulted in the discovery of containers of tungsten that were

jelled as uranium, boron poison rods in reactor fuel assemblies, and other similar anomalie

i most recent innovative use of the IVF instrumentation has been its use as a uranium

:ector. Since the derivation of Equation 2 was general in nature, it will apply equally

.1 to other gamma emitters. Since the count rate is directly proportional to both the

>tope weight fraction and density, if the former is know, the latter may be determined.

'en an appropriate gamma ray signature, uranium content can then be determined if the 23

5

U

itent is relatively invariant.

:hough 23

8

U has no useful gamma rays, it's granddaughter 234mPa does, as was discussed

rlier. If the assumption that 234mPa is in radioactive equilibrium with 238U can be made,

;n the Pa isotope can be used as a signature for uranium.

.s principle has been applied to the sorting of aged metallic uranium scrap fuel cores

:o scrap categories. In this way low yield scrap can be segregated from materials contain-

; higher uranium concentrations. The cost savings resulting from the use of the IVF

;trumentation in this way alone has more than paid for the initial investment.

'mally, because the IVF instrument is simple to operate, the day-to-day verification

tsurements are performed by a nontechnical chemical operator. Results are interpreted

I reports are issued by technical personnel. However, because of this simplicity the

';tem cannot be universally applied.

! !f

] lpletely unknown receipts, for which simplifying assumptions cannot be safely made, are

:i' rond the capability of the instrument. Difficulties also arise when portions of a ship-

" it contain gamma emitters (e.g. 137 Cs, or 60 Co) in either the matrix or containment vessel.

. the shipment is uniform with respect to extraneous gamma emitters the IVF standardization

- ; >cedures will compensate for them. If, on the other hand, the gamma spectral interferences

y within a shipment, an alternate approach such as Ge(Li) gamma spectrometry must be

i>lied

.

.ce the IVF instrumentation is stationary, field verification work is performed using the

- viously described SAM-II DCA system. Some items monitored in-situ were 15 to 18 ft. long

- 'ium-bonded , uranium carbide rods and massive uranium ingots. The stabilized, dual channel

•ay system, with direct enrichment reading capability, greatly simplifies in-situ 235U

surements. Future plans also include an evaluation of a transmission measurement for the

ification procedure to correct for matrix attenuation. This would require the acquisi-

: n of a third modular SCA system and a gamma emitter with energies resolvable using the

::!3 (Tl) detection system.

conclusion, the IVF instrumentation has proven itself to be a reliable, accurate, relative

:: fast nondestructive means of veryfying the 23

5

U content of receipts returned to the FMPC

reclamation of uranium values. The instrumentation has also found utility in the area

uranium concentration measurements for materials possessing reasonably similar 23

5

U enrich

ts. The IVF is normally calibrated and standardized once daily by technical personnel,

also supervise all nonroutine verification operations and calculate and report all

ults. Nontechnical operators perform all routine measurements obtained on a day-to-day

is.

231
TABLE I

SELECTED FMPC SNM RECEIPTS

Shipper Matrix Description

DUN U Metal Remelt scrap; Zr & Cu clad

NFS UNH Solution

ANL U0 2 S.S. & Zr clad rods

ORGDP Incinerator Ash 6" D. x 15" H. , S.S. Containers

AI UC Slugs, wafers, Ni clad plates

BAPL U Alloy Foil; Nb/Zr alloy scrap

NBL U Concentrates 5 gallon buckets

Y-12 UFi, in Paraffin Blocks, shot, wafers

SRP Filter Cake 55 gallon drums

BNWL Na2 U0,, Powder & Chunks

PAD Contaminated Scrap Vacuum bags

RMI U Residues Quench sludge, sump cakes,

extruded tubes

UNI U in concrete Saw fines & turnings in

concrete blocks

TABLE II

MEASUREMENT PARAMETERS FOR COMMON URANIUM COMPOUNDS

Metal uo 2 u 3 o 8 UO3 UF^ UF 6 "UNH"

U wt % 100 88.1 84.8 83.2 75.8 67.6 47.4

y/p(cm2 /g) 1.60 1.43 1.38 1.35 1.25 1.12 0.83

p (g/cm3) 19.0 11.0 8.30 7.29 6.70 4.68 2.81

y (cm-1) 30.5 15.6 11.4 9.87 8.34 5.26 2.33

T (mm) 0.98 1.92 2.62 3.04 3.60 5.71 12.9

TABLE III

EXAMPLE OF IVF ANALYZER CALIBRATION

Voltage: 1.2 kV;

AE: 0.30

1.0

1.1

1.2

1.3

1.4

1.5

1.6

1.7

1.8

1.9

2.0

2.1

2.2

2.3

2.4

2.5

2.6

2.7

2.8

Coarse Gain: 1/16;

Counting Period: 20 sec;

No. 1 Analyzer

76090

56298

L x + 53494

61464

73699

Pi 77331

70983

Li - 52162

38271

32740

29373

26793

25120

23392

21698

20685

18721

17668

15596

Fine Gain:

Material

:

1.6;

K-25 Oxide

L 2

No. 2 Analyzer

60671

48819

L2 "* 50513

60749

71467

P2 71271

57400

39894

32104

28155

25875

23701

22186

21012

19427

17922

16628

15246

13916

232
TABLE IV

EXAMPLE OF K-25 OXIDE STANDARDIZATION

Wt %

23

5

U

(x)

1.011

1.347

1.598

1.856

2.168

Zx = 7.980

Ex 2 = 13.535

x = 1.596

cts./3 min.

Gross

465,966

498,516

542,059

570,181

601,398

Bkg.

270,350

264,416

280,459

274,286

267,460

Zxy = 2,204,296

N = 5

Net (y)

195,616

234,100

261,600

295,895

333,938

Zy = 1,321,149

Zy 2 = 3.6057 x 10 11 y = 264,230

y = a

Q + a lX

_xy

Zx Zy

Zx 2 - Tao* 119 ' 825

ASSUMED EQUATION FORM

.2 _

y - a x x = 72,988

= 0.999

119,825

wt % 23

5

U CALIBRATION EQUATION

TABLE V

EXAMPLE STANDARDIZATION AND UNKNOWN CALCULATION FOR

RECEIPTS WITH VARYING 23

5

U AND U CONTENT

(H) High Standard (c/3m) = 167,651

(M) Intermediate Standard (c/3m) = 149,473

(L) Low Standard (c/3m) = 82,324

(c/3m) Bottom of Container = 68,420

(c/3m) Top of Container = 70,294

Vendor's % U Value = 38.68

Vendor's wt %

23

5

U Value = 1.3480

FMPC wt %

23

5

U Value (Bottom) = 1.344

FMPC wt % 235U Value (Top) = 1.381

FMPC wt %

23

5

U Value (Average) = 1.362

(FMPC - Vendor) wt % 23

5

U Value = +0.014

233
CONTAINER (CLADDING) THICKNESS

APPROXIMATE AREA VIEWED

BY COLLIMATOR

URANIUM

MATRIX MATERIAL

FIGURE 1 Detector Configuration for 235 U Enrichment Measurement

234
FIGURE 2 Void Correction Alogrithm

235
.7 DWG. 3.78

Pb COLLIMATOR THICKNESS (INCHES)

FIGURE 3 Collimator Thickness Optimization

236
0 2 4 6 8 10 12 14

NqI(TI) XTAL THICKNESS (cm)

FIGURE 4 Detector Thickness Optimization

237
238
0 20 40 60 80 100 120 140 CONTAINER THICKNESS (MILS)

FIGURE 6 Effect of Container Thickness on

23 5 U Enrichment Measurement

239
240
wnmanmD3 do NoiiDvad

241
0-

Q_

O

Q.z

CD

SAMPLE

Pb SHIELD & COLLIMATOR

INTEGRAL Nal PROBE ASSEMBLY

SCA 1 GROSS NC -11 SCA 2 BKG. nc- n

UP/DOWN COUNTER

NJ-20

RATEMETER

NR-10

SCALER TIMER NS-30A NT-29

FIGURE 9 Schematic Diagram of 235 U Enrichment Verification Instrumentation

242
Isotopic Verification Facility: (A)Conveyoi,

Accessible from Outside; (Bi Monitoring Table

(C) Nal Detector Assembly. (D) Dual Channel

Analyzer System

243
THRESHOLD POTENTIOMETER (E)

FIGURE 11 Uranium Calibration Spectra

244
FIGURE 12 Determination of Single Channel Analyzer Settings

245
FIGURE 13 Nuclear Safety Considerations for IVF Acceptance Criterion: i0n:

(A) Storage Mass, (B) Critical Mass, (C) Maximum Error in 2 35u Value for Storage Mass

246
Accurate Determination of 5-25 mg of Uranium by Redox Titrimetry

by

S. D. Reeder and J. R. Delmastro

Allied Chemical Corporation, Idaho Falls, Idaho

ABSTRACT

A precise and selective potentiometric titration method for determining 5-25mg of uranium has been developed. The method is essentially

a scaled-down version of the modified Davies-Gray titrimetric method. 1 Our investigations extend the work of Slanina et al. 2 by identifying

several critical steps where careful control of conditions is essential

to obtain high precision. The method has been automated by interfacing

a 10-mL Mettler buret with a pre-programmed KP-9830 calculator. The

calculator controls the delivery of titrant so that the end point is

approached as rapidly as possible, but delivers only small increments

of titrant near the equivalence point. A hard copy of the data (ml of

titrant vs. Electrode Potential) is printed as the titration proceeds,

and the volume of titrant equivalent to the uranium in the sample is

computed from the first derivative of the titration curve.

A precision and accuracy of 0.05% or better is obtainable with

this method. The method is especially useful for the determination of

uranium when the quantity available for analysis is limited. It has

been used for verification of the uranium concentration of various

standards used in analyses performed at the Idaho Chemical Processing

Plant.

KEYWORDS: Uranium determination; computer controlled titration;

low-level analyses

INTRODUCTION

To fill a need for an accurate independent verification of uranium concentration values

for standards and the determination of uranium in accountability samples of limited size at

the Idaho Chemical Processing Plant (ICPP), a search for a suitable analytical method was

undertaken. The range of the method was to be limited to a few milligrams of uranium and an accuracy of about + 0.05% was to be achieved with an analysis requiring a minimum of time.

Considerable experience had already been obtained by coworkers in the determination of

uranium in standards prepared for distribution throughout the world as part of the Safeguards Analytical Laboratory Evaluation (SALE) Program. One of the most widely used procedures in this program is the titrimetric determination of uranium by a method developed by

Davies and Gray 3 and modified and improved by Eberle, Lerner, Goldbeck and et al. at the New Brunswick Laboratory. Although this method has been widely accepted as precise, accurate

and selective for most nuclear fuel cycle materials, large amounts of uranium are used per

analysis (50-300 mg) and large volumes of waste solution are generated (^ 250 mL). It was

( reasoned that if this method could be scaled-down by a factor of ten and still retain the

same accuracy and precision, it would be ideally suited to the present need. As an example,

it would be practical to determine 233-U concentration in a stock solution used to spike

samples for the determination of uranium by isotope dilution mass spectrometry.

1 Eberle, A. R., Lerner, M. W., Goldbeck, C. G. and C. J. Rodden, NBL Rep. 252 (1970).

Slanina, J., Bakker, F. , and W. A. Lingerak, "Safeguarding Nuclear Materials, Vol II 1976"

p. 157.

3 W. Davies and W. Gray, UKAEA Report TRC-716(D), 1964; Talanta 11 , 1203, (1964).

247
In the modified Davies and Gray method uranium is reduced to U(IV) with Fe(II) in a phosphoric-sulfamic acid medium. Excess Fe(II) is selectively oxidized by nitric acid to Fe(III) in the presence of Mo(VI) catalyst and the uranium is subsequently oxidized to a potentiometric end-point with standard potassium dichromate solution. Vanadyl sulfate is

used to catalyze the titration and sharpen the end point. The method is quite selective a: if the sample is first fumed with sulfuric acid only silver and tin are reported to interfere. J*' 5 » 6 » 7 '

In proceeding with the development of a scaled-down method, the provision for real-ti

control of the potentiometric titration and data reduction by means of software rather tha: strictly hardware was considered to be of prime importance. Such a software controlled

system would permit maximum flexibility in performing the titration. An interface was designed and constructed by our Measurements Systems staff which permits a Hewlett-Packard

9830A programmable calculator to automatically control a Mettler DV-11 buret drive, acquir

data after measuring equilibrium electrode potentials, and handle acquired data to determi

the equivalence point in various titrations.

EQUIPMENT (CALCULATOR-CONTROLLED AUTOMATIC TITRATION)

w

:»

ill!

it

Figure 1 is a block diagram of the programmable titration system. The electromet

amplifier has an input resistance of 10 ohms and a unity-gain bandwith of 2K-Hz, and

compatible with almost any electrode system. The programmable-gain amplifier has a so ware-selected gain of 1 to 16 in steps of one to enable use of the full analog/digital

convertor resolution with any realistic electrode output. The ADC resolution is 1:409

and the full scale potential range is +2.5 to -2.5V.

PROGRAMMABLE

GAIN

AMPLIFIER

erso i ft6 an

I

itll

ilti

SAMPLE

HOLD ^5

START

CONV

ANAL06 TO

DIGITAL

CONVERTER

EOC

SEQUENCING

LOGIC

GAIN

it

DIGITAL

LATCHES

8-BIT

MULTIPLEXER

SET CTL

BURET DRIVE

TTL-I/O

INTERFACE

Figure 1. Programmable Titration System

TO CALCULATOR

A. R. Eberle and et al., NBL-252, 1970.

5 A. R. Eberle and M. W. Lerner, NBL-262, 1972.

6L. Z. Badnar, J. M. Scarborough and M. W. Lerner, NBL-265, 1972.

7 J. M. Scarborough and L. Z. Bodnar, NBL-267, 1973.

'k

248
The calculator is programmed in BASIC language. Using the proper OUTPUT statements it

direct the buret drive to deliver a specified increment of titrant, then input bits of

ry information from the ADC and convert this information to electrode potentials. Data

squired and printed for each titrant increment only after equilibrium has been attained,

is sensed by waiting a pre-selected time after addition of titrant before measuring the

trode potential and then measuring the potential again after a second time delay.

Ertial readings must agree within specified limits before a value is accepted and

ier addition of titrant is possible. With a 10 mL buret, the Mettler DV-11 is capable

slivering increments as small as 0.002 mL.

The calculator was initially programmed to add increments of titrant as manually

red from the keyboard and then print out the total volume of titrant delivered vs. trode potential. As development work progressed and the results indicated that a able analytical procedure was possible, the calculator programs were extended to allow

natic and systematic delivery of titrant, calculation of the end point (corrected for

ant temperature and buret calibration), calculation of uranium concentration, plotting

he data and data storage on magnetic tape. This hardware-software combination allows

ations to be performed to a precise end-point in a minimum of time without the

ability of introducing human errors in manipulating or reading a buret or in determining

potentiometric equivalence point.

J. Slanina et al. 2 have also reported on the use of a Mettler titrator in combination

a programmable calculator to determine small quantities of uranium by a miniaturized

i es and Gray method. They used a complete Mettler System including a DK-15 incremental

roller and apparently used the calculator only for recording and handling of data. They

titrant continually until a preset potential is sensed and then add fixed increments

each the equivalence point. We have found that the exact values of the potential prior

he end point are quite non-predictable, as they depend on the condition of the platinum

'trode and the sample solution. A program which depends entirely on expected potentials

potential changes to control the size of titrant increments was evaluated, but it did

yield high precision results.

The best precision and accuracy has been obtained with a program which adds titrant

'redetermined increments to approach the end-point systematically. To use this program

must have a close estimate of the volume of titrant required to reach the equivalence

it. Since samples requiring accurate results are run in triplicate (or more) to obtain

.istical data, the first determination is run "manually", and the results are used to

:ulate the approximate equivalence point for additional determinations on the same sample.

1 the systematic program the last 4 mL of titrant are first added in the order 2, 1, 0.5,

, 0.1, 0.05, and 0.02 mL. Increments of 0.004 mL are then added until the potential has

reased from about 500 to 630 mV. The equilibrium potentials and corresponding titrant

jmes are printed out with an HP 9866A thermal printer. The maximum AE/mL usually occurs

.T een 540 and 610 mV and approximately 5-7 minutes are required to complete an unattended

ration.

Figure 2 is a facsimile of a typical set of data imposed on a plot made with a HP 9862A

:ter controlled by the HP 9830A calculator.

EXPERIMENTAL

In miniaturizing the Davies and Gray method to determine only about 5 to 25 mg of

nium, the quantities of reagents specified in the New Brunswick procedure were initially

'.iced by a factor of ten except that 0.02N standard dichromate solution was used as a

I rant instead of 0.05N_. A 30-mL beaker was used as a vessel to carry out the preliminary

nistry and the potentiometric titration. Miniature electrodes were mounted in a polyylene stopper which nested in the top of the beaker.

Although initial results with the scale-down procedure were generally within + 0.3%

cisicn and accuracy, considerable experimentation was required to reduce this spread to

i acceptable range. One major factor contributing to poor results was the improper control

'temperature during the molybdenum-catalyzed nitrate oxidation of excess Fe(II) reductant.

i emperature of 35-40°C is generated naturally in the large scale method from the addition

' sulfamic and phosphoric acids to the sample. In this temperature range the complete

249
772r

704-

2= E3Bf

£ SB*

GAIN OFFSET STD DEV OF OFFSET

3 -13.8 0.265

VOLUME POTENTIAL 1^ (AUTO) 0.000 0.4977 1 7.440 0.6194 17 3.420 0.4363 2 7.445 0.6479 18

5.120 0.4326 3 7.455 0.6812 19 6.420 0.4310 4 7.475 0.7000 20

6.920 0.4342 5 7.515 0.7211 21 7.220 0.4424 6 7.320 0.4534 7 7.580 0.7345 22 (MAN) 7.370 0.4652 8 7.700 0.7516 23 7.390 0.4741 9 7.800 0.7598 24

7.400 0.4786 10 7.900 0.7655 25

7.420 0.4884 11 8.00 0.7724 26

7.415 0.4904 12 7.420 0.4977 13

7.425 0.5067 14 7.430 0.5229 15 7.435 0.5425 16 TEMP - 21.0

1st DERIVATIVE e.p. 7.4375 (7.4266 CORR) ML

SAMPLE SIZE - 0.56087

RESULT - 31.487

TITRRNT VDLUME(ML)

Figure 2. Typical Data Plot

destruction of excess Fe(II) occurs in a relatively short, predictable time without sin; taneous oxidation of TJ(IV) to U(VI). In scaling down the procedure the smaller quantit

of reagents do not provide and maintain the proper temperature for this reaction to pro predictably. The use of a small constant temperature water bath (35-38°C) effectively

eliminated this problem and with other parameters held constant good precision was obta

The temperature of the sample during the titration was also found to affect the re If the sample is retained in the water bath at a temperature of about 37°C during the t tion, the results are low by about 0.1%. The error is less if the sample is removed fr the water bath but an accuracy of less than + 0.05% is readily attainable If the sample

rapidly cooled to about 18°C prior to titration. Partial air oxidation of uranium at t higher temperature is believed to cause the low results. A nitrogen atmosphere can be

with a warm titration temperature but cooling the sample is preferred, since the electr

response is generally more rapid.

High results are encountered when the phosphoric acid reagent contains dichromateoxidizable impurities. This problem is effectively eliminated by dissolving enough sol

potassium dichromate in the 86% phosphoric acid tc give a persistent light yellow color

The excess dichromate is reduced with uranium by ferrous Ion and therefore presents no blem in the analysis.

PROCEDURE

The analytical procedure as presently developed is outlined below. The quantities

concentrations of reagents specified have been optimized for the pure uranyl nitrate so tions studied. Close adherence to the instructions given for each step in the procedur

given results that are better than the desired + 0.05% (typically + 0.02%) accuracy and

relative standard deviation.

250
PROCEDURE

-feigh an aliquot of the sample containing 5 to 25 mg of uranium into the

bottom of a 30-mL beaker.

Add 0.2 mL of 9M l^SO^ to the sample

and gently mix. Evaporate the sample

slowly on a hot plate until the appearance of heavy SO3 fumes and solids begin

to form in the beaker.

To the cooled sample are added 0.6 mL

water, 0.4 mL of 1.5M sulfamic acid,

5 mL of ^86% phosphoric acid, and a MD.3

rx 1.5 cm magnetic stirring bar.

Place the beaker in a small water batb

i positioned on a magnetic stirrer and

adjusted to a temperature of 35+2°C.

Stir the sample slowly for 30 s; then add

0.5 mL* of 1.4M FeSOit in 2M U2 soh directly to the sample.

Continue the gentle mixing for 1.5 min.

*Add only 0.3-0.4 mL of FeSOi^ solution

if the sample contains only 5 - 12 mg of

uranium.

j Add to the sample 1 mL of a 8M HN0 3 -0.15M

. sulfamic acid - 0.5% ammonium molybdate

. reagent. Continue gentle stirring for

; 3 min after the brown coloration cisappears.

|, Pinse down the walls of the beaker with

5 mL of LM t^SO^ containing 3 mg of

: VOSO^'R^O per mL. Without delay, remove

the sample beaker from the hot water

, bath and place it in a cold water bath

j which has a temperature of M.0°C.

:. Position the sample in the cold water

j bath under the buret and electrode

assembly and swing the magnetic stirrer

in place to support the sample for the

titration. Adjust the stirrer to mix the

sample at a moderate rate.

; Initiate the "automatic" or manual

titration program and titrate the sample

' to the equivalence point.

I

COMMENTS

For maximum precision with a 10-mL buret

select a sample aliquot containing 5 to 12 mg

of uranium when working with 0.01N I^C^Oy

and 10 to 25 mg of uranium when using 0.02N^ K2 Cr20 7 .

This step reduces all samples to a minimum

starting volume and removes volatile potential interferences such as Br~, Cl~, I-

, and

NO 3-.

The water and sulfamic acid may be used to rinse down the walls of the beaker. Do not

rinse the walls with the phosphoric acid

which has been treated with K2Cr£07 to remove oxidizable impurities.

Uranium is reduced to U(IV) leaving an excess

of Fe(II). In the absence of nitrate this

reaction will proceed without sulfamic acid.

If nitrate is present (i.e., H2S01+ fuming

step is omitted) the sulfamic acid will prevent the interfering nitric acid oxidation

of Fe(II) and U(IV) in the strong phosphoric

acid solution.

Excess Fe(II) is selectively oxidized to Fe(III). Time and temperature are critical

to this step. The temperature of the water

bath should be 37+2°C.

The sample is diluted with sulfuric acid in

this step and vanadyl sulfate is added to catalyze the oxidation of uranium with standard potassium dichromate solution and sharpen

the end-point. The solution is cooled to prevent air oxidation of uranium during the

titration step.

The buret should be filled, and the calculation programs loaded for execution.

For best results the titration should be

completed in less than 10 min. The platinum

electrode should be cleaned in cone HNO3 and

flamed in a Meeker burner after 3 or 4 determinations to assure rapid response and

a large potential change at the equivalence

point.

251
RESULTS

Following the procedure outlined previously the following set of results was obta:

using ^0.02tJ K2 Cr207. The sample was an analytical control made from 99.995% natural I containing 31.486 mg of U per gram of solution.

Weight of Sample (g) Cone. Found (U mg/g)

0.72099 (22.79 mg U) 31.487

0.63367 31.487

0.59630 31.489

0.57576 31.485

0.48628 . 31.480

0.43801 (13.79 mg U) 31.487

X6 = 31.486

S = 0.0017 rsd

S,% = < 0.01 rsd

No significant bias

Using smaller samples of the above control and diluting the K^C^Oy to 'vO.OlN, the

following set of data was obtained.

Weight of Sample (g) Cone. Found (U mg/g)

0.23706 (7.46 mg U) 31.487

0.25595 31.486

0.31935 31.497

Next day

0.26632 31.491

*0. 27189 31.476

*0. 30397 31.510

*0. 32531 (10.24 mg U) 31.484

*Samples not fumed with 112804. X7 = 31.490

S = + 0.034 % rsd

Although the results are more variable using ^O.OIN K^C^Oy as the titrant, they a quite acceptable. The samples which were not fumed with sulfuric acid exhibit the larg

variance from the mean.

CONCLUSION

The range of the Davies and Gray method for determining uranium has been successfu

reduced by about a factor cf ten without any loss in accuracy or precision. Since smal

quantities of uranium are required there is little waste of this accountable element,

titration under calculator control is considerably less tedious than manual titration,

the method is especially suited to the determination of uranium in special nuclear mate

requiring safeguarding and accountability.

252
In Situ Quantitative Determination of Transuranic Elements

in Areas of High-Level Gamma Radiation*

R. L. Brodzinski and N. A. Wogman

Battelle, Pacific Northwest Laboratory, Richland, Washington

ABSTRACT

A technique is described for passive neutron monitoring of transuranic elements. The method provides quantitative determinations of

transuranic element concentrations in a variety of field situations

where no other measurement method is possible. The technique can measure concentrations of transuranic oxides as low as 8 nCi/cm 3 and is

capable of operating in gamma radiation fields up to megarads per hour.

Information on chemical and isotopic composition can also be obtained

from the data. Several successful applications of the technique are discussed.

M

KEYWORDS: Transuranics; neutron monitoring

INTRODUCTION

The in situ detection and measurement of transuranic elements is an extremely difficult

blem. These isotopes are primarily alpha-emitting nuclides which emit relatively low

rgy gamma-rays through very low branching fractions. Transuranic elements also emit

tracteristic x-radiation in high yield, but this can only be measured in special circumnces . Neutrons are the only characteristic radiations emitted by the transuranic elements

ich are readily distinguishable in the presence of other radionuclides . These neutrons e from spontaneous fission or from (a,n) reactions on light isotopes such as

19 F, 17 0, or

For many applications, such as criticality, accountability, or waste management, re- ring the quantitative determination of plutonium or other transuranic elements, these neutns are the only radiations suitable for measurement. For example, the characterization of

iiual plutonium in process or fabrication hoods undergoing decommissioning is not possible

alpha detection and is only occasionally possible by x-ray measurements. Although gamma

ection will frequently work for fairly high quantities of plutonium, shielding by massive

hinery may introduce substantial uncertainties . Neutrons are not as severely attenuated

',bnlk quantities of steel, and remote pockets of plutonium may be detected by neutron mea-

:

, ements

.

It may be necessary to measure plutonium concentrations for potential criticality evalion in trenches, cribs, or storage tanks which have received process wastes. These locains frequently have large quantities of fission products present which would interfere

' h the detection of plutonium photons, and only neutrons can be unequivocally identified,

'dlarly, it may be desirable to measure the quantity of transuranic elements released to

environs during an accidental discharge of fission product waste or to determine the

.nsuranic element content of fuel cycle fission product effluents for disposal purposes.

I high level gamma radiation present in these cases also restricts the detection of trans-

-nic elements to neutron monitoring.

This communication describes a technique for quantitatively measuring the neutrons

.tted by transuranic elements. This technique can evaluate the neutron energy spectrum

I thereby infer the chemical and/or isotopic composition of the transuranics in question.

lis paper is based on work partially supported by Rockwell Hanford Operations' Environment

- Sciences Group, L. E. Bruns and H. H. Wang, Program Coordinators, under United States Deirtment of Energy Contract EY-76-C-O6-I83O

.

253
Finally, this technique can make quantitative measurements of transuranic element concen tions at <10 nCi/g in gamma radiation fields exceeding megarads per hours.

EXPERIMENTAL

There are two basic detection techniques for measuring neutrons emitted by transura

species. ihe first of these is the active technique, which incorporates real time measur

ment of neutrons with 3 He, BF3, fission chamber, or similar detectors and collection and' storage of signals with appropriate electronic equipment. This technique is not applica

to operation in high gamma radiation fields, since the electronic detectors will not fun

properly. In addition, neutron energy evaluation is virtually impossible with this acti

technique, since the detectors are generally operated in a moderating medium to effectiv

utilize their enhanced efficiencies at thermal neutron velocities.

The second neutron detection technique, a passive neutron activation method, is the

utilized in this work. The passive technique involves the exposure of various materials

transuranic generated neutrons and subsequent remote determination of the neutron-induce

reaction products. This technique is not influenced by high gamma fields and can yield

gy related spectral information with appropriate use of threshold reactions

.

The following criteria were considered in choosing appropriate materials for passiv

neutron monitors. The neutron-induced reaction product had to be easily determinable,

was quickly narrowed to radioactive products having distinctive gamma-ray emissions whic

could be nondestructively determined in high sensitivity gamma-ray spectrometers . The p uct had to have a relatively high branching fraction through the gamma cascade of intere

and a half-life between a few hours and a few days to allow the sample to reach near sat tion values within a reasonable exposure period. Time required for transfer from exposu

to counting precluded very short half-lives. The material needed to be in a physically

manageable and durable chemical form. The material needed to be readily available at re able cost; and cross sections for the neutron-induced reactions needed to be as large as sible in the interest of sensitivity.

Jii!

ffl

: :

a

lit m

After a thorough examination of the chart of the nuclides and a few preliminary exp

ments, the number of suitable passive neutron monitors was narrowed to the four metals,

magnesium, copper, zinc, and indium. The neutron reactions of interest in these metals

listed in Table I along with the product half-lives, gamma-rays measured, and branching

fractions. The copper and indium capture reactions monitor the thermal neutron flux and

used in different situations rather than simultaneously. The copper reaction is used in. cases where time and proximity to laboratory-based large crystal Nal(Tl) multidimensiona

gamma-ray spectrometers permits. If rapid data acquisition or remote field measurements

required, the indium reaction can be used in conjunction with a portable anticoincidence

shielded Nal(Tl) well counter at nearly comparable sensitivity levels.

The zinc reaction has a practical threshold of 2.3 MeV, and the excitation function

rises smoothly through a maximum of 280 mb at 10 MeV. The energy of the neutrons produc

by (a,n) reactions on oxygen or fluorine is dependent not only on the kinetic energy of

alpha particle but also on the excitation state of the residual nucleus. Therefore, the

reactions result in a spectrum of neutron energies with the maximum being 3.22 and 5-76

for plutonium alpha particles on fluorine and oxygen, respectively. The spontaneous fis

neutrons have a typical fission neutron energy distribution curve peaking at 1 to 2 MeV

tailing up to M.7 MeV. The zinc monitor reaction is therefore sensitive to both (a,n) a spontaneous fission neutrons

.

The magnesium monitor reaction has a practical threshold of 6.2 MeV and is, therefo

not sensitive to (a,n) neutrons but only to the small percentage of high energy spontane

fission neutrons above this threshold level. The excitation function for this reaction

smoothly through a maximum of 195 mb at 13 MeV, and although this monitor is not as sens

as the others, it provides a specific measure of spontaneously fissioning isotopes such

21t0 Pu.

254
Table I. Passive Activation Neutron Monitor Reactions

Product Branching

Reaction Half-Life Gamma-Rays Measured (MeV) Fraction

63 Cu(n,ct) 61t Cu 12.8 h 0.511 x 0.511 coincidence 18.5$

.' 15 In(n,a) 116mIn 5^.0 m 2.391 - 2.529 doublet 86. Of.

61*Zn(n,p) 61t Cu 12.8 h 0.511 x 0.511 coincidence 18.5$

2l*Mg(n,p)

2lf Na 15.00 h 1.369 x 2. 75*+ coincidence 99-85$

In order to quantitatively determine unknown transuranic element concentrations, iso-

., or chemical composition from a measurement of induced activities in passive metal

•ors, it is necessary to calibrate the technique with known quantities of transuranic

les under similar environmental conditions. This can be accomplished by constructing

tup facilities virtually identical to the system being monitored and spiked with known

ities of transuranic elements, by making rough mock-ups with "bugs" of transuranic ele-

| in strategic locations, or by monitoring an analogous field system that has been samand analyzed in the laboratory to determine its transuranic element composition. Al-

;h each of these techniques has been used successfully in this work, this paper will deal

itail with only one of the more versatile calibrations.

The major demand for this passive activation technique thus far has been for the deter-

: .ion of plutonium in soils, salt cakes, and waste tanks. Plutonium monitoring in soils

Pediments is of interest where high level fission product liquid wastes have leaked from

' ige tanks or where low level process streams have deposited accumulations over the years,

minations of plutonium concentrations in high level liquid wastes or salt cakes are of

rest from a potential criticality standpoint. These field situations all have the com-

'equirements that the monitoring must be done in situ from dry wells and in gamma-ray

"is up to megarads per hour. These requirements mandate the use of the passive activation

lique

.

' A calibration facility for the above applications has been constructed from a section of

)t (1.22 m) diameter corrugated culvert filled with local aggregate. A 6-inch (15 .k cm)

;ter dry well has been installed 11.1 cm off center of the culvert, and five 2% inch

cm) diameter source tubes are spaced on 12.7, 22. 9» 3^.9, ^8.3, and 6l.O cm centers

the dry well. Figure 1 shows a top view of this soil calibration facility. Nominal

*ram sources of plutonium oxides and fluorides of known isotopic composition are placed

ie source tubes, and the metal monitors are exposed to the neutron flux in the dry well,

ced activities are determined by gamma-ray spectrometric methods. The plutonium in each

;e tube is assumed to be uniformly distributed throughout a spherical shell of soil havi mean radius equal to the source tube-dry well separation distance. By normalizing the

:ed activities from each spherical shell to the plutonium concentration in that shell and

.ng over all significant shells, the relationship between induced activities and an "in-

*>e" source of uniformly distributed plutonium is determined. This method of calibration

)een verified experimentally for gamma radiation. A germanium diode "down well" probe

;imilarly calibrated using various gamma-ray sources, normalizing the data to each sphershell, and summing all significant shells. 1 Subsequent calibrations of this diode in a

IB h tt facility filled with uniformly spiked soil yielded identical efficiency factors,

calibration of neutron-emitting isotopes is completely analogous.

RESULTS

For the sake of brevity, only a single calibration experiment in this facility will be

-'fossed in detail. However, the results of all other calibrations performed to date are

1. At this time, only calibrations in dry soil {h.1% moisture content) have been com-

; r3d. However, it is planned to repeat these experiments at various soil moisture levels

3 saturation values . Figure 2 shows the thermal neutron flux as measured in the dry

by copper monitors for a 98.02 gram plutonium oxide source having the isotopic composiof 0.588$ 238Pu, 72.659$ 2 3 9 Pu, 19.373$ 240 Pu, 5.^27$ 21tl Pu, 1.907$ 2lt2 Pu, and 0. L Nielson, N. A. Wogman and R. L. Brodzinski, Nucl . Instrm. Methods 1U3 , 385 (1977)

255
nme

a 0°2o„°o A l

1

12.7 cm ^=g%

22.9 cm r&^'£m

34.9 cm 3feg^^

48.3 cm

61.0 cm

DnO:,

P°0°o o,Q:' 0 AGGREGATE

O^: . u.. .0 O-Oi '-

FIGURE 1. TOP VIEW OF SOIL CALIBRATION FACILITY

FOR PASSIVE TRANSURANIC NEUTRON

MONITORING SYSTEM

256
3.0

12 24 36 48 60 72 84 96

R (cm)

: IGURE 2. THERMAL NEUTRON FLUX FROM A PLUTONIUM OXIDE

SOURCE MEASURED THROUGH VARIOUS THICKNESSES

OF DRY SOIL

257
21tl Am placed in each of the source tubes. In this experiment, a significant contribut:

to the total induced activity was obtained at the maximum source-monitor separation avj

able, and it was necessary to extrapolate the data to a distance of 112 cm before the c tribution became less than 1% of the total. This extrapolation, which amounted to 20%

the total integrated value, was necessary due to the low moderating capacity of the drj

and hence the long range of the fast neutrons. Extrapolation is not anticipated to be

sary at higher moisture levels where the moderating ability of the soil will be greatly

creased. The zinc monitor required a similar extrapolation which amounted to only 1.7/

the total integrated value. No extrapolation was required for the magnesium monitor d; because very few collisions are required to reduce the neutron energy below the reactic

threshold.

For these calibrations, the monitors were all 15.2k cm diameter discs, 3.175 mm tl for copper and 6.35 mm thick for zinc and magnesium. In practice, the geometry of the

tor can be tailored to specific needs or requirements with no adverse effects. The on

necessary consideration, which has been included here, is the amount of self shielding

sented by the monitor. An experiment offered in proof incorporated the exposure of a I |j disc "facing" a plutonium source in the nearest position and another disc "edge on" to same source. Both discs yielded identical neutron fluxes which reflects the isotropic

of the multiply scattered neutrons. Self-shielding corrections are normally not requi

for zinc and magnesium monitors due to the low macroscopic cross sections for fast neu

interactions

.

Thermal neutron flux values are readily calculable from a determination of the qu

of reaction products in the monitor, a knowledge of the cross section, and application

standard radioactive growth and decay equations. The situation is somewhat more compl

the zinc and magnesium reactions, however, since the measured activities are generated

spectrum of neutron energies at widely varying cross sections. Therefore, it is most

to determine the average product of the fast neutron flux and the cross section normal

to a unit mass of monitor material. Such calibrations are determined from saturation

ity data and are reported as the number of reactions per second per gram.

fell

6

It was found that the most meaningful method of expressing plutonium concentratio

the soil, activity per unit volume, also produced the most consistent relationship bet

sources of different isotopic compositions. In the second column of Table II, the nor ized thermal neutron fluxes as measured by the copper monitors are given for one pluto

fluoride source and three plutonium oxide sources of different isotopic composition,

shown in the same column are two thermal flux values as measured by the indium monitor

verification of the reliability of the technique. Two epi thermal neutron fluxes, obta^

by enclosing the monitor discs in cadmium blankets during exposure, are also given. Is should be recognized that the thermal/epithermal ratios are related to moderating capa

(moisture level) of the soil and thus can be used to determine the appropriate set of ! bration curves in situations where soil moisture levels cannot be determined directly,

the third column, the activity normalized reaction rates for zinc and magnesium monito

are given, and in the last column, the magnesium reaction rates are normalized to the

concentration since this is the only isotope which contributes to reactions in this mo<i

Agreement between the three plutonium oxide sources is excellent and average valu

each monitor are shown in Table II. Similarly, the agreement between the oxide and fl source data is consistent with the predicted (a,n) neutron production rates in the two;, ah: pounds except in the case of the magnesium monitor data which should be independent ofp >; ical composition since only spontaneous fission neutrons contribute to this reaction,

can be seen in the table, this plutonium fluoride source yields approximately a factor

three more high energy neutrons than the oxide sources, a phenomenon which is real and

mains unexplained.

DISCUSSION

This technique has a practical detection limit of ^8 nCi/cm3 for monitoring trans; ic oxides with even lower values measurable in certain circumstances. A word of cauti: regarding background measurements is advisable here. The normal cosmic-ray neutron fl sea level is ^0.017 cm

-2 sec

-1

, and a piece of copper taken from the laboratory shelf i*

258
Table II. Neutron Reactions in Metal Monitors for Plutonium Sources

in Soil Having k.1% Moisture Content

Thermal or Epithermal Fast Neutron Fast Neutron

Flux Reactions Reactions

e and Monitor (n cm~

2 sec~

1 /nCiqcm~ 3

) (g

-1 sec VnCiqcm" 3

) (g~ 1 sec~ 1 /mg 2 'tQ Pu cm~

3

)

i {Q.8h% 2 "*°Pu)

i| Cu (2.30 ± 0.29) * 10" 3 In (2.29 ± 0.39) • 10- 3 Cu(Cd) (4.15 ± 0.61) • 10- 1*

: In(Cd) (4.53 ± 0.5M : lO" 1* Zn (5.52 ± 0.65) * 10- 8

: Mg (2.29 ± 0.65) • 10- 9 (2.11 ± 0.60) ' 10- 3

\ {5.691% 2,+ 0 Pu)

Cu (U.5 ± 1.0) -10- 5 In (3.6U ± 0.95) • 10- 5 Zn (1.03 ± 0.22) • 10" 8 Mg (7.3 ± 4.6) 10- 10 (9.8 ± 6.2) • 10- 4

(16.789% 21*°Pu)

: Cu (6.8 ± 1.4) • 10~ 5 Zn (1.00 ± 0.21) • 10- 8

[ Mg (8.8 ± 4.3) • 10" 10 (7.6 ± 3.7) • 10~ h

(19.373% 21t0 Pu)

Cu (6.0 ± 1.2) • 10-5 Zn (1.01 ± 0.22) • 10-8

Mg (6.7 ± 2.8) • IO- 10 (7.6 ± 3.2) • lO" 4

"ted Ave. PUO2

t« Cu (5.8 ± 1.5) • 10- 5 Zn (1.01 ± 0.15) * 10- 8 Mg (7.8 ± 1.7) ' 10- 10 (8. It ± 1.8) • 10- 4

-a measurable intrinsic G1*Cu activity. If this same piece of copper is kept 10 meters

X ground for a day, the 6l| Cu activity will be substantially reduced. It is therefore

fsary when making very low level transuranic element determinations to derive all back-

; d corrections based on monitors exposed under similar conditions but in transuranicnt free environments

.

i\ In certain instances, this technique can yield information on the chemical composition

tfcte transuranic elements (i.e., metal, oxide, fluoride, etc.) by the copper/zinc, copper/

sium, or zinc/magnesium ratios. Similarly, it may be possible to determine the isoai composition of a transuranic source by the magnesium/copper or magnesium/ zinc ratios.

c::This technique has been successfully applied to the quantitative determination of plu-

::m concentrations in a variety of field situations where other methods of analysis were

sible because of inaccessibility, excessive gamma radiation, or real potential for

ment contamination. Plutonium concentrations have been measured in high level fission

xfj.ct waste storage tanks, in cribs and trenches which have received plutonium-bearing

ss effluents, and in deep sediments which received accidentally released high level

on product wastes . A plutonium inventory has been performed on a building contamina-

.uring a fire as a prelude to decontamination efforts. Process and fabrication hoods and

barrels have been examined and inventoried with remarkable accuracy. In these latter

, an element of directionality was introduced by "backing" the monitors with cadmiummoderating collimators. In one waste barrel, a single slug of plutonium was located

rSth 3 cm, identified as plutonium oxide, and the weight was predicted to within 13%. In

<7-ps the most spectacular success of the technique to date, a hidden "pocket" of plu- m was found in a recess in a channel iron support beam of a fabrication hood being

259
decommissioned. The quantity and location of the plutonium was predicted, and the dai indicated it to he ^75% oxide and ^25% metal. When the hood was breeched, a pile of I oxidized plutonium metal shavings which had been generated over a period of years by i milling machine in the hood was recovered from where they had fallen through a crack s collected. The quantity recovered was ^12% greater than predicted and from with 8 cm the identified location.

260
Neutron Correlation Counting for the

Nondestructive Analysis of Nuclear Materials

by

Martin S. Zucker

Brookhaven National Laboratory, Upton, Long Island, New York

ABSTRACT

The technique of correlating the neutron pulse train from

spontaneously fissioning material as an assay tool for

nuclear material, its advantages, problems, and the underlying theory are discussed.

KEYWORDS: Spontaneous fission; plutonium nuclear material

assay; neutron correlation counting, neutron coincidence

counting

261
This report will discuss the use of neutron correlation counting as a way

of assaying spontaneously fissioning nuclear material, usually plutonium, but

also in certain situations U^-^ and other transuranic elements, particularly

Cf252 # -jhg intent is a broad rather than rigorous treatment of the theoretical

and practical aspects, summarizing work done at many institutions, principally

Brookhaven National Laboratory, the Naval Research Laboratory, Los Alamos Scientific Laboratory, and Gesellschaft fUr Kernforschung , Karlsruhe. 1 In a still

developing technique there are still some matters of opinion. In these cases we have fallen back on the developments and experiences at Brookhaven.

The basic purpose of the neutron well correlation counter is to enable the

neutrons spontaneously emitted material to be used as a valid signature for the

amount of the isotope of interest present. While basically a method for

measuring the spontaneously fissioning nuclides the amount of element present

can be inferred if the isotopic composition is known. There is currently a strong effort to develop methods to determine isotopics by gamma spectroscopic

techniques so that the neutron correlation assay method can be carried out in

the field in a timely way without recourse to laboratory mass spectrographic

analysis

.

The neutrons coming from a typical sample have three origins:

i) spontaneous fission, which depends only on the amounts and proportion of

the so-called even isotopes present in the sample (U^S, Pu^O , Cf^^,

etc.

)

ii) induced fission caused when neutrons (from whatever source) are absorbed by the isotopes present which as a result undergo fission. This

depends not only on the isotopes present, but also on their concentration

and the geometry of the sample

iii) (K,n) nuclear reactions caused by the bombardment of various (usually

light element) nuclei (present in the sample either intentionally or as impurities) by the alpha particles copiously emitted by the transuranics

nuclides present. This contribution to the neutron emission will depend on chemical composition and perhaps also the micros true ture of the material in the

sample

.

-1 At present the two most complete descriptions of the neutron correlation

method are:

R. Sher, "Operating Characteristics of Neutron Well Counters", BNL 50332,

Brookhaven National Laboratory (Long Island, New York, U.S.A.) (1972).

K. Bbhnel, "Die Plutoniumbestimmung in Kernbrennstoff en mit der Neutronenkoinzidenzmethode" (Karlsruhe, F.R.G.), KFK 2203, Gesellschaft fUr

Kernforschung M.B.H. (1975).

Both of these have extensive bibliographies to work done at BNL, LASL,

Karlsruhe, and other institutions.

262
Concentrating for definiteness on the problems of Pu measurement, of these,

ily (i) is strictly tied to the Pu isotopic content alone, and so only the

lontaneous fission can be relied on in all cases to furnish a valid indication

: the Pu content. Source (ii), induced fission neutron, cannot practically be

.stinguished from the spontaneous fission neutrons since their multiplicity of

lission and energy distribution are quite similar. Fortunately it is only for

issive and/or highly compact materials that this effect becomes appreciable,

len it does occur it is often possible to make approximate allowances, as will

; discussed later.

It is the last contributor, the so-called uncorrelated (because they are merated singly in a Poisson distributed fashion) or («,n) neutrons, which

luse most of the trouble. Their energy spectrum is sufficiently similar to

lat of fission neutrons so that discrimination on this basis is not practical,

ley will be present in any size of sample and since they depend on details of

lemical content and physical structure are usually poorly related to the amount

: Pu present.

Therefore a method using spontaneous neutron emission as a signature or idicator of the amount of Pu present must be able to eliminate the (i\,n) com- ment (and hopefully find some accommodation with the induced fission).

The basic difference between fission neutrons (spontaneous or induced) and

«(,n) neutrons which is used is the fact that while the individual fissions

ccur Poisson distributed in time, all the neutrons from a single fission are

nitted almost instantaneously (~10- ^ sec) compared to the scale of times we

re concerned with. They thus form a group in time which maintains its coherice even after moderation. The basic scheme of the correlation counter exloits this fact. It might be noted here that fissions emitting single neutrons

re of course indistinguishable from (A,n) neutrons, but the derivation of the

narac teristics of the instrument automatically takes this into account.

Before proceeding further, it is appropriate at this point to note that

nother perhaps more obvious way of exploiting the multiplicity is to use coinidence detection of emitted neutrons, and this has indeed been done. The pros

ad cons of coincidence versus correlation counting will be discussed later.

Several kinds of electronics have been developed to separate fission events

rom others producing neutrons. The simplest type just involves two scalers,

ne with a dead time of the order of the lifetime of neutrons in the detector,

he other ideally having a vanishingly small dead time. Since the first scaler

ill miss closely spaced pulses caused by neutrons from a fission whereas the

ther will record every neutron, the difference in count rate between the two

calers can be related to the number of fissions detected. There are several

aults such a scheme has, particularly that it requires elaborate corrections

jor count rate effects.

More efficient circuitry has been developed based on a correlation techique in which each incoming pulse from the neutron detector is in effect caused

o answer the question as to how many pulses preceded it in a time period of the

rder of the neutron lifetime in the detector and which therefore might have

ome from the same fission event. This type of circuit in its present form

sing shift registers to form the solid state equivalent of delay lines makes

263
efficient use of incoming information, is capable of high count rates, requires

minimal correction for count rate effects, and will be the kind discussed below.

The neutron well correlation counter consists of a detector and associated

electronics. The detector is essentially a cavity (into which samples to be

assayed are placed) formed by moderator and proportional counters sensitive

to neutrons. The purpose of the moderator is to slow the neutrons emitted

by the sample so that they may be more efficiently detected by the counters.

However, this moderation is accompanied with the introduction of a spread

in the detection times of neutrons coming from the same fission, as well

as a loss of neutrons to competing process such as leakage from the detector

or by absorption by the moderator. The effect of the moderation, capture,

and leakage, is to introduce a probability of detection per unit time which

to a good approximation can be taken to be decreasing exponentially and characterized

by a single time constant parameter known as the "die away time",t0

. This

is the neutron lifetime in the detector referred to above. Thus the probability

of detection per unit time (normalized) may be written as (1/t0 ) e"^vt0 ^

The zero of time can be considered from when the first moderated neutron

of a group from a single fission enters the detector region of the counter

after passing through the moderator.

To describe the operation of the electronics: The pulses coming from the

detector, each representing a detected neutron, are amplified and replaced on a one for one basis by uniform pulses of relatively short time duration. This

pulse train enters a pulse correlation circuit. This latter circuit analyzes

the pulse train from the detector by finding out and recording for each pulse as

it arrives, how many pulses preceded it within a particular time period, set to be of the order

The number in the register, after making allowances for counts generated by

chance (i.e. uncorrelated) pulses, can be calculated exactly in terms of the

amount of spontaneously fissioning material, q, the probability distribution

function for the number of neutrons emitted per fission P^, , the efficiency of

the detector for detecting a single neutron £ , and a factor 0 ^ F J- 1 which pertains to particulars of the pulse correlation circuit and can be calculated

knowing the details of the circuitry. In principle the calibration factor for

the neutron well correlation counter, i.e., the proportionality constant between

instrument's output the and the amount of material can be

established without using standards.

In the following the constant of proportionality will be derived with

emphasis on making the process of correlation seem physically plausable rather

than on rigor. The derivation will be made with the simplifying assumption that

events whether correlated or not (i.e. fissions or (°<,n) reactions) do not appreciably overlap in time. The conclusions would be essentially the same even

if they did. At first it will be assumed that the time over which correlations

are measured is long compared to the die away time. These conditions imply a low counting rate, but in practice, the operation is as predicted by the formula

to be derived here for count rates high enough to be at the limits of resolution

for proportional counter pulses. The breakdown of the derived relation comes about from multiplication and at extremely high count rates from electronic

problems

.

264
Consider the simplified block diagram for the correlation circuit (Figure

Each pulse entering the circuit sets off a strobe signal which activates

e adder. The pulse then enters three places; (1) a gross scaler, so called

cause it counts every incoming pulse; (2) a delay line of length 2r set by the erator in accordance with principles to be discussed later; (3) the additive

put of an up-down scaler.

Pulses exiting the delay line are routed to the subtractive input of the

-down scaler, thus the state of this scaler indicates the number of pulses in

e delay line. When the adder is activated by a new incoming pulse it thus

ds an amount equal to the number of pulses in the time preceding the incomg pulses to the register.

Suppose an isolated group of k pulses within a time period less than

proaches the correlation circuit. The first pulse activates the strobe, but

nee there is no pulse then present in the delay line, no count is added to the

igister. However, the advent of the second pulse in the group will find a lse already present on the delay line, hence a

'1' will be added to the regis-

:r . The third pulse will find two pulses preceding, and cause a "2" to be

ded to the register, etc., and so the increment to the R register by this

•oup of k pulses will be 0 + 1 + 2 +... + (k-1) = k(k-l)/2. The increment to

e G scaler will of course just be k.

The contribution to the register R at the end of a measurement time inter-

.1 t will be calculated by considering it to be made up of two contributions,

= Rf + Rrt where Rf represents the contribution due to fissions and R that

le to (<X,n) neutrons. Similarly, the gross count will be taken as G = Gf + G^.

The probability of 9 neutrons being emitted by a fission will be symbolized

' P^. This distribution has been experimentally measured and is tabulated in

iveral publications. The conditional probability that if V neutrons are

litted, k will be detected, resulting in k pulses, is P(k/v>) = (£)t^ 1 -€r)1? ~'4"^ where G. is the efficiency for the detection of a single neutron. Thus the

oduct P^Cj^U -G)*"k represents the probability that k pulses result from

fission in which neutrons were emitted. The sum over allvalues of "9 of this

roduct is then the probability that k pulses result from a fission regardless

4ote that for k^v, (£) = 0 so that the corresponding term is identically

2ro.

)

The average number of pulses resulting from a fission and added to the G

:aler is thus

265
Interchanging summations:

» K

By appendix 1(2), the sum over k can be evaluated exactly:

= £ <>>>

Letting m be the mass of material present, A the spontaneous fission rate

per unit mass, then m^t is the number of fissions during the time period. Thus

Gf = wAf.^/t—fy^K-fyt where q

= mA is the source strength for fission. There

will also be (fc,n) neutrons produced in some ratio (characteristic of the given

sample) to the fission neutron production rate. It may be assumed that there is

also a somewhat different detection efficiency for these neutrons due to their having a different energy distribution. Considering both these

factors, an effective ratio of pulses produced by (o(,n) neutrons to those

produced by fission, 0< , may be introduced such that G = Gf . Thus

Qr^Gr^Cr^ = G$ ( I +<*) = 6<^>( / +- *) ( 'ft) -

As described above, a fission which results in k pulses will add %k(k - 1) counts to the R register. The average number added to R due to the fission com- ponent per fission is therefore

From Appendix 1(3):

2

Since there were on the**average (qt) such fissions during the time interval

t, the contribution to R from this source is therefore

This calculation implicitly assumed that an indefinite amount of time was available to collect every last one of the %k(k-l) pulses. Actually

it is advantageous from the standpoint of maximizing the "signal to noise" ratio

of measurements to collect pulses only over a range of the order of the die

away time . This will be discussed later in greater detail. Also for practical

reasons having to do with the electronics, the start of the correlation time

period is delayed a small amount of time. In other words, pulses preceding the

incoming pulse within some fixed small amount of time are ignored. Thus the

incoming pulse "interogates" the delay line only over a period of time "Z 2

266
re 0^'Z±<Z'2- Since the normalized^ probability per unit time for the arrival

the fission pulses is (1/TJ,) e - ^/ LO \ the fraction of pulses which are not

t and therefore contribute to Rf is therefore

:re ti and t!^ are the beginning and end of the correlation time interval as sured from the incoming pulse. The factor F is determined by the die away

ie of the detector, f0 , and the start and stop of the time period for correla-

>n t\ and which are determined by the electronics. The quantity X\ , termed

initial delay, is built into the electronics, while £"2 is adjustable in

:ed increments ^, selected by the operator, such that 2"2 = ~l\ + n% = 'L\ + 7s.

The quantity Rf is then modified by the factor F to be

This is the contribution to R due to pulses which are correlated with one ther , i.e., arrive in groups. There is besides this the contribution to R

3m pulses randomly distributed in time. Suppose at the end of the t seconds

:ounts have accumulated in the gross scaler. This implies an average count

:e G/t. Thus at any given time there must have been on the average (G/t)f

ints in the delay line of "length"^. (This is actu-

.y a solid state realization of a delay line called a shift register.) Since

;re were by hypothesis G counts entering the detector, this is the number of

les that a quantity of counts whose average value was (G/t) was added to the

;ister. Thus

J <v=&(C-/t)r =c>(Vt)

To recover the quantity Rf which is proportional to the amount of

ssioning material we must therefore sutract R^ from the total in the register:

= < *m~ l )> g

7

- F

;re N is the so-called net count . Thus the proprotionality between the

3unt of fission in the sample (qt) and the so called net count is

<V(>; -')> ^ F

z

267
This result has been experimentally verified in several ways. By successively disconnecting detectors the data for a plot of In N vs. ln£ can be obtained

and the £^ dependence verified to within a few tenths of a percent over several

orders of magnitude of G, limited only by counting statistics.

The dependence on %<^(^-l)/^ and F has been also verified to about one percent limited only by the accuracy to which a "known" source was calibrated, and

by counting statistics.

The net count N is perhaps to be considered the primary correlation leading

to the source strength because it accomplishes the basic purpose of the instrument, namely to allow the neutrons emitted spontaneously by the sample to be

used as a signature for the amount of material present by eliminating (0(,n) and

other uncorrelated neutrons.

Another quantity can however be defined which has useful properties. If

the square of the gross count is divided by net count as defined above, the

r"i0 (&>/) .= [£^>0 + *K£t)]yj<^yH)>F(^t)

is seen to be independent of the efficiency, proportional to the amount of material (qt) , but unfortunately dependent on (#,n) production and its ratio

(including detection efficiency) to fission production.

The surprising independence with regard to efficiency is actually analogous

to the principle made use of in the absolute calibration of gamma sources involving two gammas in cascade using double coincidence counting. The above

result has been verified for neutron correlation counting to within the statistics of the counting process, for example, by deliberately changing the efficiency of the detector (disconnecting detectors) while monitoring the (G^/N)

ratio, and by other ways. (Of course, statistics do become poorer as S is decreased, keeping counting times the same.) A use for G^/N would be, for example, to eliminate the effects of the efficiency changing with respect to position within the cavity, or to render samples comparable as to Pu content which

have different amounts of absorption or moderation.

Another useful quantity is the ratio

<^ (> -•)> £ p

It depends on the efficiency, and on the (# ,n) production ratio, but not on the amount of material present. It is often possible, by considering N,G^/N,

and G/N for the same sample, to measure the (<X,n) production rate for it. This

268
t relation has been approximately verified (~10%) experimentally by

suring (0(,n) rates for materials for which the answer was known approxiely by other means.

We would now like to consider some of the practical aspects of neutron coration systems. The detector portion of the neutron correlation counter can

made in many forms and sizes to accommodate different measurement problems.

largest to date is one to assay ~ 200 liter waste drums, the smallest we w of assayed single metal clad Pu "coupons" which were about 5 cm X 5 cm X

cm. Special ones have been made to assay 1 - 2 kg Pu buttons, and for fuel

s. The typical so called neutron well counter type illustrates features comto all (Figure 2). There is a central cavity for the sample surrounded by

ayer of polyethylene moderator followed by one or more rings of proportional

nters either He^ or BF3 filled in more moderator. The thickness of moderator

ween the tubes and cavity is determined by the following consideration. As

I iis increased, there will be a greater moderating effect on the fast neutrons

,or 2 MeV average energy) coming from fission, hence a greater sensitivity of

detectors to the neutrons because of the 1/E (where E is the neutron energy)

avior of the capture cross section of or He^ at low energies. However,

greater the thickness of intervening moderator, the more chance for neutron

ture by the polyethylene or scattering out of the reach of the detectors. As esult of these competing processes there is a maximum in the sensitivity of counter versus the thickness of moderator between the detector tubes and the

'pie cavity. This is at about 25 mm for Pu and 30 mm for Cf^52 when gp-j

led tubes are used, the difference being the greater average energy of Cf^52

trons . By making the thickness slightly less than that for maximum sensitiv-

("undermoderated") then if the sample has hydrogenous material in it there

1 be relatively little change in the sensitivity since the operating point of counter will be shifted about only in the vicinity of the maximum. Well

nters for general purposes are usually designed with this in mind either

ng a computer code or direct experimentation. If the counter is to be used

moderator free materials,

y. The combination of a

i /E propc

es typically neutron well counters which have an efficiency (20 - 30)% using

and (30 - 40)% using He^ tubes.

: j then it could be designed for maximum efficiency

1/E proportional counter together with moderator

The choice between He^ and BF3 can often be decided on the basis of cost ectiveness. He^ 2.5 cm D. 4 atm. tubes, roughly speaking, will give about

greater sensitivity than the 5 cm D 0.9 atm. BF3 tubes (these are the opti1 parameters for safeguards use), but at about a three times greater price.

cost of proportional counters is the major item in a typical neutron corre-

:ion apparatus, even including the electronics. BF3 proportional counters are

s sensitive to gamma rays and produce faster rise time pulses. They have

>rer resolution and it is suspected a shorter lifetime (though it is still

-sured in years). For nearly all applications BF3 tubes suffice.

External to the tube region there is a layer of ^.8 mm Cd followed by an

:er layer of polyethylene. This outer layer of polyethylene slows down neu-

>ns coming from outside the counter so that they can be more readily captured

the Cd and prevented from entering the counter proper. The sample cavity

lally has a ~ .8 mm Cd lining to prevent thermalized neutrons from re-entering

! cavity where they could induce fissions in the sample.

269
The typical electrical arrangements of the detector and following electronics is illustrated in Figure 3. Proportional counters are run with the center wire positive with respect to the shell. While conventionally the shell is grounded, there are advantages to having it at a negative high voltage. When

this is done the center wire will be essentially at ground voltage and can be

fed directly into the charge sensitive preamplifier without the intervention of

a decoupling resister and blocking capacitor, which are noise producing

elements. Since the tube shells are embedded in polyethylene, an excellent insulator, there is no danger from having the shells at a high voltage.

The tubes are divided into groups, each group with its own preamplifier

located right in the high voltage junction box. A multi-coax cable connects

each preamp with a following amplifier and a fast output discriminator. The

amplifier puts out a relatively slow gaussian pulse , 2^/^sec FWHM, or a differentiated gaussian, designed to optimize the signal to noise ratio for the

type of proportional counters used. This slow pulse is replaced by a standard

0.1 or O^^sec pulse from the discriminator. The discriminator pulses are combined together in an "OR" circuit which forms the pulse train analyzed by the

following correlation circuitry.

Dividing the proportional counters into groups as described, each followed

by its own preamp, amplifier, and discriminator accomplishes two objectives. The

input capacitance per preamplifier is kept small for better signal to noise

ratio. More important, the "amplifier dead time" caused by the possibility of

pulses from different proportional counters overlapping is effectively divided

by a factor almost equal to the number of groups that the counters are divided

into, yet with no loss in pulse height resolution. This resolution is important

in that it enables the discriminator bias to be set optimally between the noise

level and the smallest neutron produced pulses occurring with appreciable frequency. A measure of the excellent signal to noise ratio achieved is the fact

that proportional counter tubes designed by the manufacturer to operate at voltages of 2500 to 2800 volts with ordinary amplifiers, can be operated routinely

at 1000 - 1500 V, and even give useful signals at as low as 500 V, although a portion of the pulse shapes then deteriorate because of poor electron collection. Lower tube operating voltages have practical consequences in reducing the

incidence of high voltage leakage problems.

A problem with all well counters is that of variations in sensitivity

within the counting volume. A small source of neutrons moved about within the

volume will be counted with different efficiency depending on position; this is in fact a practical way to map out the efficiency as a function of location. For

a cylindrical volume surrounded by proportional counters whose ative length is about twice that of the cavity height the fall off in sensitivity towards the

ends of the counting volume can be about (10 - 15)% of the maximum it has at the

midpoint of the axis. (The large overhang was for the purpose of minimizing

this problem. ) This fall off in sensitivity can be lessened somewhat by using

"reflector" end plugs made of aluminum or graphite instead of polyethylene. In

a radial direction along the midplane of the counter, the sensitivity will rise

as the walls of the counter are approached, by roughly 10% more than the value

on the axis. It isn't clear that anything in the way of design improvements

could be made to alleviate this without compromising other aspects of the well

counter design, such as azimuthal symmetry.

270
Both problems can be minimized by confining the dimensions of samples

^asured to be a small fraction of the volume, and making sure that samples are

jr.tered within the counting volume. In practice these are often not possible,

le typical well counter is already of the order of 500-600 kg in weight when

ailt to accommodate 1-2 liter packages, and the lack of industry standards for

ontainers guarantees there will be some oddly sized containers to be measured.

One approach to this problem is to develop a correction factor based on a apping of the counting volume sensitivity. Another method which will work if

le (iX,n) contribution can be assumed constant is to use the (G^/N) method

escribed above. Finally, effort should be made to standardize packaging or to

3 cautious in comparing sample assay results where different geometries are avolved

.

An objection to the correlation circuit shown above (Figure 1),

ore theoretical than practical, is that it assumes that the non-correlated

ount is constant during the measurement so that it can be calculated as ^t/t. If this is not so, for example, caused by changes in the background

ount of uncorrelated neutrons entering the counter from nuclear material

eing moved about in the vicinity of the measurement apparatus, the fluctuations

a background will give rise to a correlation, and the estimate G^-tJt would

e too small. This effect can be taken into account by monitoring the instantaneous

alue of the correlation due to fluctuations in the uncorrelated neutrons

y means of some additional circuitry as shown in Figure 4.

The basic action is the same as the previous but with an addition. The

trobe signal activates the adder as before, but is also delayed for a period of

ime long compared to the delay line (correlation time)f, yet small compared to

he time over which the background might be expected to change. Since f is of

he order ( 10-100

)

/{sec , but it would take ~(1 - 10) sec for changes to be made

n the room inventory arrangements, such a choice can be made without diffiulty, say n -~10 . The delayed strobe is thus allowed to activate the adder

gain via the OR logic circuit at a time such that the pulses in the delay line

can have no relation to the incoming pulse from the pulse train, yet close

nough in time to accurately sample the instantaneous value of the background,

he prompt interrogation results are routed to a register Rp which is the same

.s the previous R, while the delayed interrogation results are routed to a regster Rj by the action of the AND circuits. Register 1 accumulates the count

4ue to (i) correlated and (ii) uncorrelated neutrons, and ( iii) } includes a con- ribution due to fluctuations in the uncorrelated neutron count rate, while regster 2 accumulates just the latter two contributions, (ii) and (iii). Thus the

.ifference Rp - R^ = N, the correlation count due to correlated neutrons.

Under normal circumstances where the background is constant it is prefertble to use the formula N = R - G^Tf/t in preference to N = R^ - R^ since its

:tatistical accuracy is better. This stems from R being subject to greater

.luctuations than G which is generally a very large number, which being a simple

';ross count, is a Poisson statistic.

When more than one isotope is present, the ^(f-l)^ must be replaced by an ippropriate average taking into account the P,> for the particular isotope

information available from e.g. E.K. Hyde, "The Nuclear Properties of the Heavy

Elements", Vol. II, pp. 221-222 and others), the fraction of all the spontaneous

271
fission nuclides that it contains , and the decay rate for fission for that

nuclide. For perhaps most samples only the Pu^O content need be considered to give a fairly good answer. People do loosely speak of the Pu^O equivalent in terms of the other two isotopes but this is only in terms of neutron yield and

would not be proper strictly speaking for ^7(^-1)^ since there are slight differences in Pp for the different even Pu isotopes. For Cf252 ? ^e are very different from that of the Pu isotopes

.

The extension of the formula for the net count to the case where there is more than one contributing spontaneous fissioning nuclide present is straightforward

A/= ± L F t JT <>>(<? >>. e£V^t A i //^

c

where L is Avogadro 1 s number (no. of nuclei per g. nuclear mass), i refers to the nuclide species of mass m^ (g), nuclear mass A^Cg), and decay constant for

spontaneous f ission /\j ( sec

-

-'- ) . If the total mass is m and the isotopic mass fraction is C. = ) then

A/=J-LFt<mY<*Ci^S> €t£At//\i

2. z—i Usually the efficiency €j can be considered the same^i.e., the neutron

energy spectra are similar, for all species present (a good assumption if for

example we were dealing only with Pu isotopes); then we can write the net correlation count per unit mass and time as - (iLFe*)(^t, -o> f, A. /Ad

L The dimensionless first factor on the right is a property only of the instrument, the second factor having units (sec~^8~^) depends on the nature of the

material. For example, a particular neutron correlation apparatus intended for

mixed oxide fuel rod assay had (1/2 LF 2) = 3.14 X 10^2 f while some fuel rods

made up of normal U with a few percent Pu had a value for the summation of^~ =

2 . 19 X 10~23 sec

- -'-? -

-'- . Such calculations are necessary for example if the

standard has a different isotopic mix than the samples being assayed.

272
Some values of parameters for Pu and U are cited below for example only,

hey are not necessarily the best values; we have an ongoing request with

he National Nuclear Data Center at Brookhaven National Laboratory to update

he required data for the nuclides of interest.

TABLE I

Calculation of r rom P ^

Pu238 Pu240 Pu242

y>(r-l) :p_ T>(?-1)PV -i>(v-i)pv

0 0 .044 0 .041 0 .063 0

l 0 .175 0 .219 0 .192 0

2 2 .384 .768 .351 .702 .351 .702

3 6 .237 1 .422 .241 1.446 .324 1 .944

4 12 .124 1.488 .127 1.524 .033 .396

5 30 .036 .720 .020 .400 .036 .720

6 60 .001 .030

r-1))> - 4.398, 4. 102, 3.762

A value of ^Y(V-1)> for U

238 can be inferred from quoted values of Diven's

Darameter Ds <V (-^-l p/<i>2 to be <Y (V-l)^ - 3.212.

TABLE II

(A is the fission decay constant on a per nuc leon, not mass, basis)

isotope <V(y>-l)> X( sec" 1) A(g) > <^(->M)>/A

Pu238 4.398 4.481 X 10"•19 238.05 8.279 X 10~ 21

Pu 240 4.102 1.877 X 10"•19 240 . 06 3.207 X 10" 21

Pu242 3.762 3.138 X 10"•19 242.06 4.877 X 10"21

U238 3.212 2.197 X 10"•24 238.05 2.935 X 10-26

273
TABLE I

Typical Values for a

fractional

element or composition

isotope element

Pu 2.83

Pu238

Pu240

Pu242

U 85.32

U238

I

Mixed Oxide Fuel Rod

(10-2) f A<*(?-l)>/A

isotope (sec

- lg- l

506 1 186 X 10"24

19 709 1 789 X 10"23

1 766 2 437 X 10"•24

99 29 2 486 X 10"26

L.- 2 154 X 10"23

It is also possible and usually more practical to calibrate the instrument

with a single piece of material of known isotopic composition, or unknown composition if the weight is known and the isotopics are the same as the samples to be assayed.

When samples are large and or dense multiplication will occur as evidenced

by an increase in the net count per gram of material over what it would be for

smaller samples of the same type. The basic criterion is whether the geometry

and macroscopic neutron cross section of the sample are such that neutrons arising from the sample have appreciable probability of reacting further with it.

While people have spoken in terms of the amount of material beyond which one gets multiplication, actually in theory there is no threshold, and whether multiplication is observed or not depends on the precision with which the assy is carried out. As a practical matter one might expect to see significant multiplication in powdered Pu02 weighing the order of 1 kg or so, while for Pu in

metallic form (e.g. in the form of alloy plates or coupons with their flat areas

in contact), a few hundred grams would suffice.

Attempts have been made to calculate the amount of multiplication in a sam274
pie using methods of reactor theory. One approach based on a simple idealized

model of the sample in the counter geometry gives the result in the form of a complicated algebraic formula of uncertain accuracy when applied to an actual

situation. Another approach using a Monte Carlo technique is capable of accuracy limited in principle only by computer size and running time but would seem

impractical to apply to field use. Perhaps the answer is preparation using

either approach of extensive tables covering a wide range of cases which the

operator in the field can use to interpolate from to fit the actual cases

met

.

A practical way to counteract multiplication recognizes that it arises

mainly from neutrons whose cross section for inducing fission was increased by

slowing down in the well counter body, re-entering that cavity and interacting

with the sample. This can be alleviated to an extent by lining the cavity with

Cd or B^O (borax or boron carbide in some convenient way of containing the

powder). The neutrons emitted in fission will pass through relatively

unattenuated , but there will be a much higher attenuation for the backward

diffusing thermal and epi-thermal neutrons. The attentuation of fast neutrons

from the sample will be appreciable, perhaps 20% for ~.8 mm Cd, with

corresponding loss in the efficiency £ but this may still be a worthwhile sacrifice. The Cd lining can be made removeable for a slight advantage when small

samples are being assayed.

The next line of attack with respect to multiplication is to accept that

the calibration curve will depart from linearity and calibrate the system for

samples of a given type by using a series of increasing weights as standards.

The geometry should be kept as comparable as possible; for example, the container diameter might be kept the same while the depth of the material within is

allowed to change.

What will be found is that the calibration curve will be initially linear,

starting from the origin but will then depart at some point in the direction of

giving a larger net count per unit weight of material. Neutron correlation well

counter measurements have been made with samples of from 1 - 2 kg of Pu02 with

pu240/pu ~ o . 1 , where the multiplication, i.e. ratio of neutrons actually

produced to those coming from spontaneous fission, have been as great as ~1.2

or 1.3. Such samples can have count rates in excess of 10^ s -

^ . Samples larger

than that, say — 3 kg, have to be avoided anyway because they would constitute

, a criticality hazard if the neutron well counter should become flooded with

water. The maximum allowable amount of Pu in metallic form would be considerably less from the criticality standpoint.

In neutron correlation measurements as in any other kind of measurement it

is desirable to maximize the signal to noise ratio. In the present situation

the "signal" are the correlated counts while the "noise" is made up of

uncorrelated counts, due to (i*.,n) neutrons, neutrons from the environment, and

I also neutrons coming from fissions such that only one of the produced neutrons

was detected. The correlation circuitry discriminates against uncorrelated

events by calculating correlation over a period of time set to be of the order

of the die away time of the detector. Typically the adjustment allowed by the

circuitry is from 16 to 128 flsec . The detector die away time is a good first

approximation as to what correlation time to use. This may be measured using

the reactor physics technique called the Rossi- X measurement . In it the pulse

275
train from the detector is used as input for a multichannel scaler. Assuming

the MCS is not busy when a pulse enters, a time sweep is set off. The next

pulse coming in during the sweep causes a count to be added to the channel the

sweep is momentarily at. The result of this process repeated many times for a detector exposed to a source of correlated and uncorrelated neutrons will be a histogram as shown in Figure 5 consisting of a simple exponential atop a flat

base. This can be interpreted as a plot of the relative probability per unit

time of the time interval between the pulse initiating the sweep and the next

following pulse. The exponential part is proportional to the probability density of detecting correlated neutrons with the detector and the decay constant

of the exponential may be identified with the die away time of that detector.

This is actually a practical way of measuring the die away time. It is also

possible to calculate die away times with a Monte Carlo or other code, but the

experimental way is probably more reliable. The flat portion of the histogram

is proportional to the probability density of uncorrelated events. The action

of the correlation circuitry can be interpreted in terms of measuring the total area between the vertical lines defining the correlation time interval and

subtracting from it that portion consisting of just the uncorrelated contribution. This difference, the correlated count, divided by the uncorrelated count,

is the signal to noise ratio for this type of measurement. Maximizing it thus

depends on the relative sizes of the two areas. For small random contributions,

one can increase the correlation time interval, for large random contributions

it should be made smaller, in order to achieve optimum signal to noise ratio.

From this discussion it follows that an optimum counter design is one which

maximizes the efficiency and minimizes the die away time. These goals are only

partially at odds with one another, since if neutrons are "lost" to the detector

volume by being absorbed by the detectors, resulting in counts, then both goals

are being served. Reducing the die away time by introducing neutron poisons (as

has been done) such as Cd sheeting is generally a poor practice, since the detection of correlated counts depends on the square of the efficiency, £p- , whereas the uncorrelated counts will be detected as C alone. Moreover in

introducing the poison one is converting some good events (i.e., more than one detected neutron from a fission) into random events (only one detected) so the

signal to noise decreases precipitously. Thus, poisoning the detector should

only be used for count rates the electronics cannot otherwise handle.

The statistics of the correlation problem have been studied by various of

the groups working in this field but with results which are not convincing to all the parties. One thing which is clear is that the correlation count is not

a Poisson statistic and therefore the theoretical standard deviation is not given simply by the square root of the number of counts. Some groups however

still use what amounts to a Poisson approach to calculating the statistical uncertainty, namely J{ corr . + uncorr . ) + uncorr.', i.e., the uncorrelated count

is treated as a background to be subtracted from the total (correlated + uncorrelated), with each being respectively a Poisson statistic. Other formulae

have been proposed, but this writer is skeptical and thinks at the moment the

best approach is to experimentally determine what standard deviations might be

expected in a given situation by repeated measurements.

We should consider correlation counting in comparison with other methods of

assaying Pu^O and similar nuclides. The use of correlation counting as a measure of fissioning neutrons arose from the decision to use B^f-j or He^ propor276
,ial counters to make up the detectors. While the original reason for using

i detectors mainly may simply have been that they were the classic neutron

:ctor, it is also true that they both have the virtue of being quite insensi-

; to the gamma rays which are always associated with plutonium and the other

isuranics. Both these counters depend upon the cross section of and He^

neutrons varying (for low enough energies) essentially as 1/E. To make

;e detectors have high efficiency, the neutrons have to be slowed down

lerated) by something like polyethylene. As mentioned before, this causes

zatistical spread in the arrival time of neutrons at the detectors. While

irence is maintained over times (which depend on the geometry and materials

che detector assembly) which are of the order of (10 - 100) ji sec, this

=ad renders the usual coincidence techniques unuseable for such a detector

;e the accidental rate would be too great for such poor resolution. Thus,

auto-correlation technique in this situation becomes necessary.

Applying the coincidence method requires dividing the detector into

halves to be put into coincidence with one another. Hence, the efficiency

the detection of single neutrons is also halved and the coincidence efficiency

uced to a fourth of what it would be if the detector of the same bulk

= used in an auto-correlation mode. Organic scintillator-photomultiplier

e detectors would have to be used since proportional counters unznoderated would b< insensitive. The detector would consist of a plastic or liquid scintillator

ch operates either on the recoil principle or by slowing down the neutrons

'il they can be captured and cause the disintegration of some elements

0, Gd, Cd) with which the scintillator is loaded. Organic scintillators

sensitive to gamma rays. To some extent, the random gamma rays emitted

the plutonium can be discriminated aainst by the resolving time of the

ncidence circuitry, while those gammas which are emitted in the fission

cess (an average of 'v-'7.5 per fission) will give rise to valid coincidence

•nts . However, reliance on gammas will give rise to increased matrix dependence

assay results compared to the case where the neutrons alone are used because

y are more readily absorbed. Electronic neutron-gamma discrimination

emes can be used but would add greatly to the complexity of the circuitry,

ractical problem is the tendency of photomultiplier tubes to drift with

iperature and count rate, while proportional counters are notably stable

er these conditions

.

While more can be said on the question of using coincidence rather than

relation counting for the purpose of assaying Pu by utilizing the spontaneous

;sion of the even isotopes, we feel that in summation there is no good

son to develop organic scintillator-based coincidence schemes. There are worthwhile advantages to offset many disadvantages.

Active NDA equipment (random driver, ISAF etc. used in the passive mode

:h no neutron source) has been used to assay plutonium, but this is not the

>t way. The count rates tend to be low, the calibration relation is nonlear and matrix dependent and there is no longer the possibility of tracing

i basic instrument calibration back to a fundamental relation as is the case

;h the neutron well correlation counter.

277
APPENDIX I

Suppose one consider the sum (a + b), where a and b are any two numbers,

raised to a power -y . From the binomial expansion formula

fit",;- -=^_ ; a^ij

k

i > \ r

and in particular , \

( ; x

, .= c i> • *> I v )" / Take the derivative of both sides with respect to the parameter a:

> ('• < 'A- " =^k(i)^ fe '''i)' R

(o

Multiplying through by a:

k

Consider the special case a + b = 1, and let a =6, then:

k

Taking the derivative of (1) again with respect to a:

Multiplying through by a^ :

Again for the special case a + b = 1, a = (/:

278
o

I—z3oo

CO

COo

cce>

Ll_ oK3

c_I-O 3oo

(T

UJH

CO

LlI CCu_o3

Q_HO

UJ

_l

DE _J

CM to

UJ

ffl O

DCH

co

(3CO

O

co±

_l<

3tr

CHtuwI—I

Qa -p 0

(0^ c

Cn

O

H +J T3O M

rH

O I

l-l

Cn •H

279
SECT. AA

POLY. SHIELD

COVER

AND

REFLECTOR

PROPORTIONAL

COUNTER

TUBES

COUNTER BODY

(POLY.)

Cd. LINER

REFLECTOR

HV IN

TO MULTIPLE

AMPLIFIER

MODULE

PREAMPS

Figure 2 - Typical neutron wel

counter constructic

features

230
OJ

u c o O -Pu - :u

0) 3 -P Ul

(!J Q)

U

C T) -HCO

>-i

O -P -P •H C O

>i

O

H iuo

tn -H

Cm oCD

a. u_ IS

z. Q= O LJ

cn huj o 2 O

281
o

o

282
a 3

i to •H 0) (0 S-l CO0 -P a c

1 i E

nj -h O U

•H CI) a cx

E-i <U

I

CDU

CP •H

PL,

Q_UJ

LlI

CO

li_ O

CO

O

UJX

3WI1 UNO U3d

Ainiavaoad 3Aiiv"i3U

3iUJ h- Q

283
An Instrument for Monitoring the Transuranic Content

of Chopped Leached Hulls from Spent Nuclear Fuel Elements

by

N . A. WOGMAN, R. L. BRODZINSKI, and D. P. BROWN

Battelle-Northwest Laboratories, Richland, Washington

ABSTRACT

A leached hull monitor featuring high sensitivity for direct transuranic element analysis of various kinds of reactor fuels has been designed for the Commercial Nuclear Waste Vitrification Program (CNWVP) at Battelle-Northwest Laboratories. The instrument can detect the presence

of transuranics in undissolved fuel embedded in the chopped hulls following a leaching process, or in undissolved fines or other liquid or solid effluents associated with the CNWVP chemistry. The monitor detects

neutrons arising from spontaneous fission and from (a,n) reactions on oxygen in the hulls and wastes . The system is constructed using a massive external neutron shield, an internal lead gamma-ray shield between

the sample and the neutron detectors, and an electronic system which re- cords all single and coincidence neutron events which occur during a preset time interval. Both the transuranic neutron flux and the cosmic-ray

produced neutron background are determined simultaneously. The system is

calibrated with known sources of uranium and plutonium oxides and metals

with isotopic compositions typical of each reactor fuel to be processed.

The estimated detection limit of the system is 2 mg of Pu in a 10 ^ sec- ond counting period.

KEYWORDS: Design of hull monitor; plutonium; nuclear fuel; nuclear

waste

INTRODUCTION

Transuranic accountability requirements present many difficult problems in the analysij

of effluents and solid wastes during reprocessing of spent nuclear fuels. Transuranic re- siduals may be embedded in the cladding material or be in the form of undissolved fuel following the leaching process due to entrapment in the hulls by the chopping process . If the#:.

concentration of these transuranics exceeds some preset allowable limit, the hulls must be jj releached or placed in retrievable storage instead of some less expensive waste disposal

There are a variety of methods to detect transuranic elements in chopped leached

hulls or other wastes associated with the nuclear fuel recovery process. Active methods

such as the detection of delayed neutrons from neutron-induced fission, detection of neutrons from spontaneous fission, or detection of (a,n) neutrons produced by the decay of thv

transuranic elements would be the most direct means of measurement. Other less direct mea- surement techniques, which have been considered by workers in the field, are the detection

of fission or activation products which hopefully can be related to the quantity of residus

fuel. These measurements are generally made using an analysis of gamma-rays emitted by fis sion products. 1 The latter technology does not directly determine the actual transuranic

species and thus is subject to error when preferential leaching of fission products or trai: uranic elements occurs in a process stream or in the chopped leached hulls. In addition, ; high energy fission product recoils embed themselves in the cladding in such a fashion that

they may be unleachable with respect to the transuranic species, which creates an overestimate of transuranic element content. In either case, an erroneous value for the transuran:

content will be obtained by the indirect method.

L T. Gozani, "Leached Hull Monitor for Detection of Undissolved Irradiated Nuclear Fuel,"

INTEL-RT 2311-003, Report by I.R.T. Corp., P.O. Box 80817, San Diego, CA

284

(,
This communication describes a neutron detection system for the direct measurement of

e transuranic elements in the hulls of chopped leached fuel elements and some associated

[fluent wastes

.

INSTRUMENTAL DESIGN

The most direct means of measuring the transuranic elements in chopped leached nuclear

.el hulls and wastes is to detect their characteristic neutron emission. Since hulls and

.stes are highly radioactive, the neutron detection system must be able to operate in the esence of high gamma-ray fields. Neutrons can normally best be detected by 3 He counters,

3 counters, or fission chambers. However, for this application, fission detectors are too

efficient, and 3 He counters are too sensitive to gamma radiation. Hence, BF3 counters

.ve been chosen.

The monitor has been designed to accept a 5-gallon can of leached hulls or other wastes

[ich will be highly radioactive. A 10- cm thick 800 kg lead shield is located between the

japle cavity and an annular detector array as shown in Figure 1. This thickness has been

.rerimentally determined to be an optimum trade-off between gamma-ray attenuation and loss

' geometry/efficiency for neutron detection, and is expected to reduce the radiation dose

. the BF3 tubes from about 10

1* to about 7 R/hr. Also, the BF3 neutron detectors are

.ielded from excessive external gamma radiation and from external neutron sources by a mod-

.ar tank ranging from 15- to 30-cm thick filled with a saturated aqueous solution of boric

.id. The water acts as a bulk shield for gamma radiation and is a moderator for fast neu-

•ons . Once thermalized, extraneous neutrons are eliminated with high efficiency by the

>ric acid. Neutrons emitted by the residual transuranics in the sample chamber pass

irough the lead shield and enter the counter/moderator module, which consists of one hun-

-ed 5-cm diameter, 65-cm long carbon-coated BF3 proportional counters (Reuter Stokes, Inc.)

arrounded by a minimum of 2. 5-cm of polyethylene which acts as a thermalizer for the neurons . The carbon coating reduces the degradation that would normally occur from gamma-rays

le stainless steel clad tubes are filled to 90 cm Hg pressure with BF 3 which has been en-

.ched to 9G% 10 B. All tubes are matched in gain at a common point on their plateau voltage

3 ±3$ . Plateau voltages are 1000 volts long starting at approximately 1500 volts with a

.ope of 'VO. 3^/100 volts. All tubes use an alumina ceramic insulation for moisture protec-

.on and have FWHM resolutions ranging between 16.8 and 26.9%. Tube sensitivity averages

f.8 cps/NV with some as high as 60 cps/NV. All tube testing was done with a PuBe neutron

Jurce

.

Neutrons detected by the BF3 tubes generate pulses which are stored in a series of ten

; -:alers, one of which records single events. The others record double, triple, quadruple,

;c . coincidence events as shown in Figure 2. A variable time-delay window is used to de-

--.ne coincidence. Pulses are stored for counting periods ranging from 100 to 10,000 seconds

crter appropriate background corrections are made, the net scaler readings are used to esti-

! ute the quantities of transuranics remaining in the hulls

.

The remaining electronics consist of eight charge sensitive preamplifiers, four preamp-

,Lfier power supplies, two summing amplifiers, one single channel analyzer, one timer, one

tgh voltage power supply, and two NIM bin/power supplies, all of which are commercially

/ailable

.

DISCUSSION

Cosmogenic background is anticipated to be the major problem in measuring neutrons emit

;=sd by the transuranic elements. In an experiment conducted at Los Alamos Scientific Labora

pry 2

, a substantial background which was attributed to cosmic-ray initiated neutrons in a Passive lead shield containing the neutron detection equipment, was reduced by a factor of

D when the system was covered by 0.9 meter thick concrete overburden. The backgrounds for

•-ach system will thus vary with its relationship to massive shielding in each process facility.

m

It should be pointed out here that the background is not necessarily constant since the

ajority of extraneous pulses are from temporally fluctuating cosmic particle neutron pro due

ion in the massive lead shield around the central cavity. However, these events are largeT. W. Crane and H. 0. Menlove, "Application of the Shuffler System with Shielding for Small

Hot Samples," LA-6788-PR (1976)

285
ly multiple coincidence events, whereas the transuranic-produced neutrons are virtually e; clusively single or double coincidence events. The multiple scaler counting system allo\i|

rather precise definition of the single and double coincidence events of cosmic origin tc! inferred from the measured higher order events. In this fashion, the necessary backgrour;

correcting data are obtained at the same time the actual measurement is being made. Quarj

tative system calibration is accomplished with transuranic sources of varying size, disti

bution, and chemical and isotopic composition.

The system discussed in this paper has been designed and engineered but constructior

!

has not yet been completed so operational data are lacking. However, experiments which v

conducted during the design phase determined most of the operational parameters and the ] tical detection limits for the transuranic elements . Although the hull monitor can be ce brated for any type of nuclear fuel, sensitivity calculations were based on a fuel where

of plutonium emits 200 n/sec. The overall geometrical efficiency of all BF 3 tubes is Vfl ' Using an estimated counting efficiency of 27% and knowing that the lead absorber reduces

neutron flux three-fold yields a total efficiency for single neutrons of ^7% and for doul coincidence neutrons of Background for the system is estimated to be 3 c/sec for

j single events and 0.0^ c/sec for double coincident pulses based on mock-up experiments

.

background will, of course, vary as a function of overburden available to reduce the cosiij ray flux, as well as temporal variations in the cosmic-ray activity. These figures yield

detection limit of ^2 mg plutonium in a counting period of 10

1* seconds. This is at leas -

]

. factor of 10 lower than that obtainable by using non-direct photon analysis of fission

products to infer fissile material in the hull.

The system described in this paper will be used in the nuclear fuel recycle program IE? it is deemed necessary to measure the transuranic content of hulls and other wastes at tM,

10 nCi/g level.

r

id

9

286
LIFTING EYES

FIGURE 1. SCHEMATIC OF MONITOR FOR TRANSURANIC ELEMENT ANALYSIS OF NUCLEAR FUEL HULLS AND WASTES

287
NPUT

GATE WIDTH

CONTROL GATE 1 AND —¥ COUNTER

GATE 2 CONTROL

1

SIGNAL 4

CONDITIONER GATE 2 AND COUNTER

GATE 3 CONTROL — 2

' +

GATE 3 — COUNTER AND

GATE 4 CONTROL 1 1 1

3 —

—

>

i ,

1 >

r

GATE n-1 — AND COUNTER

GATE n CONTROL

n-1

* — COUNTER

GATE n n

INITIAL CONDITIONS:

FIRST PULSE:

SECOND PULSE: (WITHIN GATE TIME) THIRD PULSE: (WITHIN GATE TIME)

nth PULSE: (WITHIN GATE TIME) ADDITIONAL PULSES: (WITHIN GATE TIME)

GATE 1 OPEN, ALL OTHER GATES CLOSED COUNTS IN COUNTER 1, TRIGGERS GATE WIDTH CONTROL, CLOSES GATE 1, AND OPENS GATE 2 COUNTS IN COUNTER 2, CLOSES GATE 2 AND OPENS GATE 3

COUNTS IN COUNTER 3, CLOSES GATE 3 AND OPENS NEXT

GATE COUNTS IN COUNTER n

COUNT IN COUNTER n

AT THE END OF GATE TIME, INITIAL CONDITIONS ARE RE-ESTABLISHED

FIGURE 2. BLOCK DIAGRAM OF COINCIDENCE ELECTRONIC UNIT

288

tl U.S. GOVERNMENT PRINTING OFFICE : 1978 O—281-067 (295)
4A (REV. 5-78)

.S. DEPT OF COMM. IL'OGRAPHIC DATA

SHEET

1. PUBLICATION OR REPORT NO. NBS SP-528

2. Gov't Accession

No.

3. Recipient's Accession No.

T-E AND SUBTITLE

lytical Methods for Safeguards and Accountability

isurements of Special Nuclear Materials

5. Publication Date November 1978

6. Performing Organization Code

THOR(S)H. Thomas Yolken and John E. Bullard, editors

8. Performing Organ. Report No.

R FOR MING ORGANIZATION NAME AND ADDRESS

NATIONAL BUREAU OF STANDARDS

DEPARTMENT OF COMMERCE

WASHINGTON, D.C. 20234

10. Project/Task/Work Unit No.

11. Contract/Grant No.

loflsoring Organization Name and Complete Address (Srreef, City, State, ZIP)

lear Fuel Cycle Division, Virginia Section, American Nuclear

iety; and National Measurement Laboratory, National Bureau

Standards

13. Type of Report & Period Covered

14. Sponsoring Agency Code

PPLEMENTARY NOTES

Library of Congress Catalog Card Number 78-600115

JSTRACT (A 200-word or less factual summary of most significant information. If document includes a significant

bliography or literature survey, mention it here.) This book contains proceedings of the America

y the National Bureau of Standards) topical meeti

afeguards and Accountability Measurements of Spec

as held in Williamsburg, Va. on May 15-17, 1978.

ere to discuss the latest techniques for chemical

nd to strengthen lines of communication among sci

The presentations deal with advanced analytic

-ray spectrometry for elemental content, neutron

iques for isotopic content, mathematical correcti

or elemental content. Examples of remote systems

amples for analysis are given. Progress is being

ith obtaining high precision and accuracy for ana

hi 1 e at the same time maintaining short turn arou

iversion. Various approaches are described which

perating systems with built-in safeguards and qua

rends in the fuel cycle, along with future approa

f special nuclear materials, are discussed.

n Nuclear Society's (co-sponsored

ng entitled, "Analytical Methods for

ial Nuclear Materials. " The meeting

The two objectives of the meeting

analysis of special nuclear materia

entists working in this field,

al chemistry techniques such as interrogation and gamma ray techon models, and wet chemistry methods

for handling highly radioactive

made on the problems associated

lysis of special nuclear materials

nd time and reduced possibility of

emphasize computer-controlled

lity assurance programs. Current

ches to control and accountability

Is

-Y WORDS (six to twelve entries; alphabetical order; capitalize only the first letter of the first key word unless a proper

me; separated by semicolons

)

uclear safeguards; accountability; x-ray, gamma-ray; analysis; special nuclear

aterials; neutron.

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