NBS Measurement Services:

wiiiv iVe^ »f'p

Activation Foil

Irradiation by

Reactor Cavity Fission Sources

NBS

Special

Publication

250-14

George P. Lamaze

James A. Grundl

U.S. Department of Commerce

National Bureau of Standards
Center for Radiation Research

The Center for Radiation Research is a major component of the National Measurement Laboratory in the National Bureau of Standards. The Center provides the Nation with standards and measurement services for ionizing

radiation and for ultraviolet, visible, and infrared radiation; coordinates and furnishes essential support

to the National Measurement Support Systemfor ionizing radiation; conducts research in radiation related

fields to develop improved radiation measurement methodology; and generates, compiles, and critically

evaluates data to meet major national needs. The Center consists of five Divisions and one Group.

Atomic and Plasma Radiation Division

Carries out basic theoretical and experimental research into the • Atomic Spectroscopy ~

spectroscopic and radiative properties of atoms and highly ionized • Atomic Radiation Data

species; develops well-defined atomic radiation sources as radiometric • Plasma Radiation

or wavelength standards; develops new measurement techniques and

methods for spectral analysis and plasma properties; and collects,

compiles, and critically evaluates spectroscopic data. The Division

consists of the following Groups:

Radiation Physics Division

Provides the central national basis for the measurement of far ultra- • Far UV Physics

violet, soft x-ray, and electron radiation; develops and disseminates • Electron Physics

radiation standards, measurement services, and data for for these radia- • Photon Physics

tions; conducts theoretical and experimental research with electron,

laser, ultraviolet, and soft x-ray radiation for measurement applications; determines fundamental mechanisms of electron and photon interactions with matter; and develops advanced electron- and photon-based

measurement techiques. The Division consists of the following Groups:

Radiometric Physics Division

Provides national measurement standards and support services for ultra- • Spectral Radiometry

violet, visible, and infrared radiation; provides standards dissemination • Spectrophotometry

and measurement quality assurance services; conducts research in optical • Radiometric Measurement Services

radiation, pyrometry, photometry, and quantim radiometry; and develops

spectroradiometric and spectrophotometric standards and calibration

procedures. The Division consists of the following Groups:

Radiation Source and Instrumentation Division

Develops, operates, and improves major NBS radiation facilities

Including the electron Linac and race track microtron; develops,

designs, and builds electronic and mechanical instr unentation for

radiation programs and facilities; provides national leadership

in the standardization of nuclear instrunentation; and develops

new competence in radiation sources and instrunentation. The

Division consists of the following Groups:

• Accelerator Research

• Linac Operations

• Electronic Instrunentation

• Mechanical Instrunentation

Ionizing Radiation Division

Provides primary national standards, measurement services, and basic •

data for applications of ionizing radiation; develops new methods of • chemical and physical dosimetry; conducts theoretical and experimental •

research on the fundamental physical and chemical interactions of

ionizing radiation with matter; provides essential standards and • measurement support services to the National Measurement Support •

System for Ionizing Radiation; and develops and operates radiation •

sources needed to provide primary radiation standards, fields, and •

well-characterized beams of radiation for research on radiation •

Interactions and for development of measurement methods. The Division

consists of the following Office and Groups:

Office of Radiation Measurement

Radiation Theory

Radiation Chemistry and Chemical

Dosimetry

Neutron Measurements and Research Neutron Dosimetry

Radioactivity

X-Ray Physics

Dosimetry

Nuclear Physics Group

Engages in forefront research in nuclear and elementary particle physics;

performs highly accurate measurements and theoretical analyses which

probe the structure of nuclear matter; and improves the quantitative

understanding of physical processes that underlie measurement science.
NBS MEASUREMENT SERVICES:

ACTIVATION FOIL IRRADIATION BY

REACTOR CAVITY FISSION SOURCES

George P. Lamaze

James A. Grundl

Center for Radiation Research

National Measurement Laboratory

National Bureau of Standards

Gaithersburg, MD 20899

U.S. DEPARTMENT OF COMMERCE, C. William Verity, Secretary

NATIONAL BUREAU OF STANDARDS, Ernest Ambler, Director

Issued March 1988
Library of Congress Catalog Card Number: 88-600513

National Bureau of Standards Special Publication 250-14

Natl. Bur. Stand. (U.S.), Spec. Publ. 250-14, 56 pages (Mar. 1988)

CODEN: XNBSAV

U.S. GOVERNMENT PRINTING OFFICE

WASHINGTON: 1988

For sale by the Superintendent of Documents, U.S. Government Printing Office, Washington, DC 20402-9325
PREFACE

The calibration and related measurement services of the National Bureau

of Standards are intended to assist the makers and users of precision

measuring instruments in achieving the highest possible levels of

accuracy, quality, and productivity. NBS offers over 300 different

calibration, special test, and measurement assurance services. These

services allow customers to directly link their measurement systems to

measurement systems and standards maintained by NBS. These services

are offered to the public and private organizations alike. They are described in NBS Special Publication (SP) 250, NBS Calibration Services

Users Guide .

The Users Guide is being supplemented by a number of special

publications (designated as the "SP 250 Series") that provide a detailed description of the important features of specific NBS

calibration services. These documents provide a description of the:

(1) specifications for the service; (2) design philosophy and theory;

(3) NBS measurement system; (4) NBS operational procedures; (5)

assessment of measurement uncertainty including random and systematic

errors and an error budget; and (6) internal quality control procedures

used by NBS. These documents will present more detail than can be

given in an NBS calibration report, or than is generally allowed in

articles in scientific journals. In the past NBS has published such

information in a variety of ways. This series will help make this type

of information more readily available to the user.

This document (SP 250-14), NBS Measurements Services: Activation Foil

Irradiation by Reactor Cavity Fission Sources, by G. P. Lamaze and

J. A. Grundl, is the fourteenth to be published in this new series of

special publications. The scope of the service, the neutron field

characteristics, the irradiation procedures, and the uncertainties in

the reported neutron fluences or fluence rates are described (see test

number 44090C in the SP 250 Users Guide). Inquiries concerning the

technical content of this document or the specifications for these

services should be directed to the authors or one of the technical

contacts cited in SP 250.

The Center for Radiation Research (CRR) is in the process of publishing

21 documents in this SP 250 series, covering all of the calibration

services offered by CRR. A complete listing of these documents can be

found inside the back cover.

NBS would welcome suggestions on how publications such as these might

be made more useful. Suggestions are also welcome concerning the need

for new calibration services, special tests, and measurement assurance

programs.

Joe D. Simmons Chris E. Kuyatt

Acting Chief Director

Measurement Services Center for Radiation Research

i i i
ABSTRACT

This document describes the National Bureau of Standards (NBS)

calibration service 44090C (formerly 8,1R) which operates in the

following way: (i) dosimetry sensors (metal foils, nuclear track

detectors, wires, crystals, etc., supplied by the customer or by

NBS) are irradiated to a measured neutron fluence in a fissionspectrum neutron field (for short lived reaction products, the

samples are irradiated to near saturation and the fluence rate

specified); and (ii) the irradiated dosimetry sensors are shipped

to the customer, followed by a test report stating the fluences or fluence rates and associated uncertainties. The scope and

philosophy of the service, the neutron field characteristics,

irradiation procedures, and the uncertainties in the reported

neutron fluences or fluence rates are discussed. Typical maximum

fluences are of the order of 3 x 10^^ neutrons/cm^, with combined

(la) uncertainties of ± 2%.

Key words: activation; dosimetry; fast neutrons; fission spectrum; neutron

f 1 uence
TABLE OF CONTENTS

1. DESCRIPTION OF SERVICE 1 1.1. Confi yuration and Characteristics 2

2. NEUTRON FIELD CHARACTERIZATION 4

2.1. Neutron Fluence 4

2.2, Neutron Spectrum 6 2.2.1. Calculation 6

2.2.2. Measurement 6 2.2.3. Evaluated Spectrum 6 (multigroup format, spectrum errors)

3. IRRADIATION PROCEDURES 9

4. DETAILED OPERATING PROCEDURES 10

4.1. Assembly and Irradiation of Sensors 10

4.2. Fluence and Fluence Rate Calculations 11

4.3. Summary of Uncertainties 12

5. BENCHMARK REFERENCING 14

6. SAFETY 15

REFERENCES 16

APPENDIX A. Typical detector assembly worksheet ..... A-1

APPENDIX B. Typical irradiation log and fluence calculation worksheet . B-1

APPENDIX C. Operating instructions for gamma spectrometer pulse

height analyzer C-1

APPENDIX D. Spectrometer data analysis program - STP2D8 D-1

APPENDIX E. Scattering correction program - SCATFO E-1

APPENDIX F. Saturation factor program - SAT260 F-1

APPENDIX G. Typical Test Report G-1

V
LIST OF TABLES

Page

TABLE 1. Evaluated Fission Neutron Spectra for 252Qf gpfj 235u -jn

45-Group Format 7

TABLE 2. Uncertainty Estimates for 252cf and 235u Evaluated Fission

Neutron Spectra 9

TABLE 3. Assignment of Uncertainties 13

vi
LIST OF FIGURES

Page

Fig. 1 235u Cavity fission source 3

Fig. 2 The NBS evaluation of the ^ssy fission neutron spectrum 8

vi i
iII1jji 1I
1. DESCRIPTION OF SERVICE

The National Bureau of Standards (NBS) maintains standard fission neutron

fields (or benchmark fields) to supply calibration irradiations and certified

fluence standards. The NBS Cavity Fission Neutron Source provides a 235|j

thermal -neutron-i nduced fission neutron field spectrum. Known fluences are based upon the known source strength for the National Standard Ra-Be Neutron

Source, NBS-I [1]. The NBS calibration service 44090C [2] (formerly 8.1R)

operates in the following way: (i) dosimetry sensors (metal foils, nuclear

track detectors, wires, crystals, etc., supplied by the customer or by NBS) are irradiated to a measured neutron fluence in a fission-spectrum neutron

field (for short lived reaction products, the samples are irradiated to near saturation and the fluence rate specified), and (ii) the irradiated dosimetry

sensors are shipped to the customer, followed by a test report stating the

fluences and associated uncertainties. Typical maximum fluences are of the

order of 3 x lO^^ neutrons/cm^, with combined (la) uncertainties of ± 2%.

Dosimetry sensors must fit within a cylindrical cadmium enclosure of

20-mm-inside diameter and 6.5-mm-i nside height.

For a typical activation detector, the customer measures the activity of

the irradiated sensor with his own detector system and corrects the activity

to obtain the activity at the end of the irradiation (EOI). This activity is

then corrected for scattering effects (see section 2.1) to obtain the freefield activity at EOI. For simple decay, this free-field activity of the

neutron fluence standard in disintegrations per second is related to the

free-field neutron fluence by the activation equation:

A = X • C • N • a • $ (1)

where: A = free-field activity (Bq) at EOI,

X = decay constant (s"i) of the reaction product,

C = decay correction factor (dimensi onl ess) given in the test report

(this corrects for activity lost during the irradiation),

N = number of atoms in the target isotope,

a = fission-spectrum-averaged reaction cross section (cm^),

and $ = free-field fission neutron fluence (n/cm^) given in the test

report.

The commonly employed reaction rate probability (experimentally, the

saturated specific activity) is given by

= a . (j, (2) XCTN

1
where: (A/aCTN) = reaction rate probability (disintegrations per

second/nucl eus. )

(|) = ($/T), the effective neutron fluence rate or flux density

(n/cm2 s) , and

T = the length of the irradiation (s) given in the test report.

1.1 CONFIGURATION AND CHARACTERISTICS [3], [4], [5]

The NBS cavity fission source operates at the center of a 30 cm diameter

spherical cavity in the NBS Research Reactor graphite thermal column. The

source-detector capsule consists of two coaxial source disks of ^ssy metal

(16 mm dia x 0.13 mm thick) placed outside a cadmium box that encloses passive

detectors for exposure. The cavity and source-detector arrangement are shown

in Fig. 1. The source-detector assembly is designed so that the cadmium box

and detectors may be removed separately from the fission disks. Fission

neutron fluence rates of ~ 4 x lO^o n/cm^ s are obtained at the midplane

between the source disks when they are at a separation distance of 11 mm. For

thin disk-shaped detectors with diameters up to 13 mm, axial fluence gradients

are mild: the fluence at 1 mm from the midplane is 4% greater than at the

midplane. Center-to-edge fluence ratios are more severe — see Section 2.1.

Fission neutron return for spherical cavities in graphite has been

studied extensively by means of neutron transport calculations and by experiment in connection with the general development of standard neutron fields in

a spherical geometry [6]-[10]. For the 30-cm cavity at NBS, the response of

threshold fission detectors to fission neutrons returning from the cavity

walls is less than 0.3% of the uncollided fission neutron fluence from the

fission source disks; for detectors with thresholds above 1 MeV, the response

to the cavity return fluence is less than 0.1%. These estimates, based on calculation, have been checked experimentally by means of fission chamber

traverses in the cavity fission sources at Center for Nuclear Studies in Mol

,

Belgium (SCK/CEN) [9]. More important is fission neutron scattering in the

source-detector capsule. The scattered neutrons have a different energy

distribution then the original spectrum and therefore will generally have a different interaction probability. Monte Carlo calculations are employed for

this problem and they provide corrections for threshold detectors which vary

with detection threshold and with the number and arrangement of detector disks

[3]. Corrections for activation rates are generally between 0.5% and 3% (see

section 2.1).

The NBS fission source is not generally used for low-energy integral

detectors because of uncertainties in the graphite return background which has

a near-l/E distribution of neutron fluence. The graphite return correction

for the 235u(n,f) reaction, for example, is nearly 10% and self-shielding of

the cavity return fluence by the fission disks and by the ^ssy activation foil

makes this correction uncertain to a factor of about one third. The one meter

cavity arrangement at the SCK/CEN is better suited for low-energy response

detectors [9].

2
NBS CAVITY FISSION SOURCE

Fia 1 ^^^U Cavity fission source. Upper view: relative positions

of neutron sensor foils, fission disks and cadmium enclosure. Lower

the assembly in the thermal column of tne NBS reactor.

V 1 ew

3
2. NEUTRON FIELD CHARACTERIZATION

The energy spectra of fission neutrons are very similar for all fissionable nuclides and both test and power reactors exhibit fission-spectrum-like

components in the MeV energy range. The 235u fission neutron spectrum may be

described in terms of a broad energy range,

lower bound median upper bound

Ep(p=0.99) Ep(p=0.5) Ep(p=0.01)

235U 0.08 MeV 1.57 MeV 7.2 MeV

where p = fraction of neutron spectrum above Ep; and a coarse seven-group flux

display, (j)(E)AE, as follows:

E limits: 0 0.25 0.8 1.5 2.3 3.7 8 12 MeV

(|)(E)dE: 0.054 0.197 0.229 0.195 0.192 0.127 0.006

A detailed description of the spectrum is given in Section 2.2.

Although the neutron spectrum listed is specifically for thermal -neutroninduced 235u fission, the fission spectrum for fast-neutron-induced fission is

little different. The change in median energy is about 0.05 MeV.

2.1 NEUTRON FLUENCE (See also Sections 1.1. and 3.)

The traceability path of NBS neutron field strengths begins with NBS-I,

the U.S. national fast neutron source standard [4]. This source consists of a beryllium sphere, 4.0 cm in diameter, at the center of which is placed one gram of radium in the form of RaBr^. The radium is enclosed in platinumiridium capsules of 0.2 mm wall thickness. This source has been calibrated by

several independent methods [11-13] and the source strength (n/s into 4ir steradians) is now established as 1.243 x 10^ ± 0.85% (la). The NBS 252cf

spontaneous fission neutron source strengths are obtained by measuring their

ratio to NBS-I in a manganese sulfate bath [14], The uncertainty of the source strength obtained in this fashion is ± 1.1% (la). The 235u fission

neutron fluence in the cavity is established by means of fluence transfer from

the 252Qf fission neutron irradiation facility. The neutron fluence at the

midplane of the source-detector capsule for a source disk separation of 12 mm

is about 0.12 n/cm^ for each neutron emitted from the fission disks which

corresponds to a fluence rate of ~ 3.5 x lO^^^ n/cm^ s with the reactor at

20 megawatts full power. The fission neutron fluence rate is minimum at the

midplane between the two source disks. At 1 mm above or below the midplane,

the mean fluence for a 12 mm diameter disk detector is ~ 4% higher than at the

midplane. Radial variations are more severe: center-to-edge fluence ratios

are 1.2 for 12 mm diameter disks. This gradient does not require that

detector disks be extremely uniform. A radial mass nonuniformi ty of 20%, for

example, distributed linearly in a detector foil changes the activation by

less than 0.5% compared to a uniform mass distribution. Also, the nickel

fluence monitor included in each irradiation package yields a fluence averaged

over the radial variation.

4
The fission neutron fluence is perturbed by a small cavity return contribution, and also by scattering in fission source disks, in encapsulations, and

in the detector foils. The correction for these effects, based on calculation, is given as the fractional departure for each detector from its

free-field activity:

total

measured = activity

f ree-f iel

activity 1 * ^c]

This correction depends markedly upon reaction threshold because of the change

in neutron energy caused by scattering in the nearby materials of the

encapsulation. Components of y^^ , obtained from detailed Monte-Carlo

calculations [15] of the source-detector capsule with a representative foil

stack, are as follows:

cavity return

encapsulations

(incl. source disks and cd box)

foil stack

y5(,(%)

change in

neutron

f 1 uence

+ 1.2%

+2.5%

+1.5%

changes in threshold detector response

for the indicated changes in neutron fluences

E95 = 1.2 MeV = 2.1 MeV = 6.5 MeV

+(0.17± 0.1)% < 0.1% < 0.1%

+(0.5 ± 0.3)% -(0.2 ± 0.3)% -(1.3 ± 0.3)%

+(0.0 ± 0.4)% -(0.6 ± 0.4)% -(0.8 ± 0.4)%

+(0.7 ± 0.5)% -(0.8 ± 0.5)% -(2.1 ± 0.5)%

The uncertainties listed are la. The spread of the foil -stack component of

is associated with foil position. It is possible to derive from the Monte

Carlo calculation individual foil-to-foil scattering factors as a function of

foil separation. These correction functions can be applied to a range of foil

stack arrangements beyond the one calculated as long as the foil-to-foil

scattering is mainly first collision. An overall uncertainty of ± 0.5% (la)

is assigned to y^^. This uncertainty has been estimated from the distribution

of the scattering corrections from a large number of foil positions in several

different Monte Carlo calculations.

It is the small source-to-detector distance which makes it possible to

obtain nearly pure fission neutron fluences in small cavities. The cavity

return fluence at the cavity center is about 1.2% of the uncoil ided fission

neutron fluence. This return fluence gives rise to threshold detector

responses that are between 0.05% and 0.3% of the free-field fission neutron

response depending upon the reaction threshold.

5
2.2. NEUTRON SPECTRUM

2.2.1. Calculation . Fission neutron spectra are most accurately determined by laboratory measurement because all the calculations have undetermined

fitting parameters. (See, however, Ref. [16] and earlier work cited

therein.)

2.2.2. Measurement [17], [18], From 1952 to the present, wel 1

documented measurements of 252cf and ^ssy fission neutron spectra have

employed every perfected technique of neutron spectrometry [19]. Accordingly,

fission spectra are the most widely studied fast-neutron energy distributions

in existence. The large body of documented fission spectrum data was evaluated at NBS in 1975 [18]. The evaluation included an estimate of the

spectrum uncertainties in a multigroup format based on the spread of the

various data sets. The 252cf ^^id ^ssy fission spectrum shapes recommended in

the evaluation are based on the sixteen selected measurements listed in

Table 3 of ref. [20].

2.2.3. Evaluated Spectrum . The NBS fission spectrum evaluation has

been chosen for presentation in this document because of the consistency of

its procedures for both 235u ^^^^ 252cf fission spectra, and because it

includes an articulated uncertainty estimate [17], [18]. Many other

evaluations are available.

The NBS evaluated spectrum is described up to 20 MeV by means of a reference Maxwellian, M(E), modified by four piecewise continuous linear

segments below 6 MeV and one exponential segment above 6 MeV. The reference

Maxwellian for the ^ssy fission spectrum is,

M(E) = 0.750 /r exp(-1.5E/1.97) , E in MeV,

and the adjustment functions y(E) are as follows (here 25 denotes ^ssy).

Energy Interval (MeV) ^25^^)

0 - 0.25 1 + 0.8E - 0.153

0.25 - 0.8 1 - 0.14E + 0.082

0.8 - 1.5 1 + 0.040E - 0.062

1.5 - 6.0 1 + O.OIE - 0.017

6.0 - 20 1.043 exp [-0.06(E - 6.0)/1.043]

The evaluated spectrum is given by x(E) = y(E) • M(E). A 45-group

tabulation of the evaluated ^ssy fission spectrum is given in table 1 and

plotted in figure 2. The evaluated 252cf fission neutron spectrum is given

for comparison purposes. Group fluences for other energy group structures may

be derived from this tabulation with the interpolation function recommended in

the footnote to the table. Spectrum uncertainties, given in table 2, are based on standard deviations from the mean of subsets of measured data from

the final adjusted Maxwellian of the evaluation. This estimate, carried out

in a seven-group structure, is based on the scatter of the data from the

sixteen experiments which were evaluated.

6
TABLE 1. EVALUATED FISSION NEUTRON SPECTRA FOR 252cf AND 235u

IN 45-GROUP FORMAT

252cf

Lower

Enerav

Boundary

(MeV)

Group

Fl uence

(() • aE

Cumul ati ve(

Fl uence to Lower

Bounda ry

Group

Fl uence

(j) . aE

Cumul ati ve^^)

Fl uence to Lower

Boundary

0.0 0.0039 1 .0000 0.0048 1 .0000

0.05 0.0074 0.9961 0.0088 0.9952

0.1 0.0219 0.9887 0.0249 0.9864

0,9 0.0140 0.9668 0.0154 0.9615

0.25 0.01 52 0.9528 0.0166 0.9461

0.3 0.0323 0.9377 0.0350 0.9295

0.4 0.0337 0.9054 0.0363 0.8945

0.5 0.0343 0.8717 0.0367 0.8582

0.6 0.0343 0.8374 0.0365 0.8215

0.7 0.0338 0.8031 0.0358 0.7851

0.8 0.0664 0.7693 0.0697 0.7493

1 .0 0.0641 0.7029 0.0668 0.6796

1 .2 0.0608 0.6388 0.0629 0.6128

1 .4 0.0290 0.5780 0.0298 0.5499

1 .5 0.0279 0.5490 0.0286 0.5201

1 .6 0.0527 0.5211 0.0536 0.4915

1 .8 0.0484 0.4684 0.0487 0.4379

2.0 0.0442 0.4200 0.0441 0.3892

0.0206 0.3758 0.0204 0.3451

2.3 0.0196 0.3552 0.0193 0.3247

2.4 0.0364 0.3356 0.0356 0.3054

2.6 0.0328 0.2992 0.0319 0.2698

2.8 0.0296 0.2664 0.0284 0.2379

3.0 0.0503 0.2368 0.0477 0.2095

3.4 0.0310 0.1 865 0.0290 0.1618

3.7 0.041

0.1555 0.0378 0.1328

4-2 0.0253 0.1 1 42 0.0227 0.09503

4.6 0.0200 0.0889 0.0176 0.07233

5.0 0.01 90 0.0689 0.0164 0.05473

5.5 0.0140 0.0499 0.0118 0.03833

6.0 0.01 02 0.0359 0.00831 0.02653

6.5 0.00734 0.0257 0.00573 0.01822

7.0 0.00527 0.01839 0.00394 0.01249

7.5 0.00378 0.01312 0.00271 0.00855

8.0 0.00270 0.00934 0.00186 0.00584

8.5 0.00193 0.00664 0.00127 0.00398

9.0 0.001 37 0 .004/

U .UUUo

/

n n n 9 7

U m\i\iC.I 1 9.5 0.00098 0.00334 0.00059 0.00184

10.0 0.00118 0.00236 0.00068 0.00125

11 0.00059 0.00118 0.00031 0.00057

12 0.00030 0.00059 0.00014 0.00026

13 0.00015 0.00029 0.00007 0.00012

14 0.00011 0.00014 0.00004 0.00005

16 0.00003 0.00003 0.00001 0.00001

18 0.00000 0.00000 0.00000 0.00000

For interpolation up to

a = 0.70 for 252cf; a = 10 MeV, use the shape

0.76 for 235U.

function E^ . exp(-aE);

2 35U

^^^Spectrum fraction above lower energy bound.

7
8
TABLE 2. UNCERTAINTY ESTIMATES FOR 252cf and 235u EVALUATED

FISSION NEUTRON SPECTRA

2 35U

(Spontaneous Fi ssi on) (Thermal -Neutron- Induced Fission)

Energy

Boundari es Group

Fl uence

Uncertai nty

la

(%)

Uncertai nty

Group la

Fluence (%)

^ • AE <|) • AE

0.0

0.25

0.8

1.5

2.3

3.7

8

12

20

0.047

0.184

±13

± 1.1

n nc /I

0.197

4-1 C

± 4.1

0.220

0.194

0.200

0.146

± 1.8

± 1.0

± 2.0

± 2.1

0.229

0.195

0.192

0.127

± 3.0

± 3.1

± 2.0

± 4.8

0.0087

(0.00058)*

± 8.5 0.0056

(0.00026)*

± 5.3

Above 12 MeV, there is insufficient data to obtain accurate fluence

estimates.

Of the many alternative descriptions employed for the ^ssy fission

spectrum, the most common is the Watt-function fit chosen for ENDF/B-V [21] in

June, 1979:

.j;(E) = C exp(-E/a) sinh(/bE)

a = 0.988 MeV b = 2.249 MeV"! C = 0.4395

median energy: Ep(p=0.5) = 1.64 MeV ,

Contemporary time-of-fl ight data were emphasized in the ENDF/B-V evaluation as compared with the NBS evaluation which considered all wel 1 -documented data.

3. IRRADIATION PROCEDURES:

Disk detector irradiations are monitored with nickel or indium monitor

foils placed between the detectors. The neutron fluence in the detector

capsule is obtained from the measured ^QNi(n,p) or ^^5ln(n,n') activity on the

basis of a fluence transfer calibration carried out at the NBS 252cf

Irradiation Facility, In this procedure, the ratio of the activity obtained

in a known 252Qf fission neutron fluence to that obtained during a cavity

fission source irradiation is translated into a ^asy fission neutron fluence

at each monitor foil position in the source-detector capsule.

9
The expression which governs this fluence transfer is

_ ^

fcf_ . $ = ^ . $

cvy D

- cf

^cf ""cvy

G ^ a -:(>E ) \1) ^{>E ) D

= cf cf^ ^cf^ p^ cvy

G

' o (>E ) *

ijj (>E ) * D . cf

cvy cvy^ p' xvy^ p' cf

where $ = fission neutron fluence and "cvy" denotes the ^ssy cavity fission

source and "cf" the 252Qf fission neutron calibration fluence,

R = the reaction probability, and

Bp = truncation energy (p = 0.95 commonly).

The ratios on the right side of the equation involve decay correction

factors (G), truncated reaction cross sections a(>Ep), spectrum fractions

i|;(>E ), and the ratio of the monitor foil activity, D, in the cavity to that

at the 252cf irradiation facility. Details of the fluence transfer procedure

are given in Refs. [4], [22]. These references also describe a set of

validation experiments, carried out at the SCK/CEN Facility, in which

23^Pu(n,f), with negligible fission spectrum shape sensitivity, was effectively substituted for 5SNi(n,p) and ii5ln(n,n') as the fluence transfer

reaction. The fluence transfer procedure yields a free-field 235y cavity

fission neutron fluence to an accuracy of ± 2.0% (la).

Fission neutron fluences of ~ 3 x lO^^ n/cm^ may be obtained at the NBS

facility in 24 hour long irradiations. The small irradiation volume restricts

multiple foil packages to a diameter of 12 mm and a thickness of a few millimeters. The gamma-ray dose from neutron capture in graphite (~ 0.7 kGy/h) and

cadmium (~ 4.4 kGy/h) is about 46% of the neutron and gamma-ray dose from the

fission sources.

4. DETAILED OPERATING PROCEDURES

4.1. ASSEMBLY AND IRRADIATION OF SENSORS

4.1.1. If the sensors are unmarked, they should be marked with a permanent identifier at this time, preferably by scribing. Sensors should be

cleaned with acetone and weighed after drying. The weighing accuracy should

be 0.2% or better. Masses should be recorded on the worksheet and the plastic

bag containing the sensor. The masses of the nickel and/or indium foils are necessary to calculate the fluence; other masses are measured to obtain a complete record of the radiation.

4.1.2. The sensors should be mounted into the detector capsule. The

positions of the sensors should be recorded on the worksheet (see appendix A)

and in the notebook.

10
4.1.3. Arrange time at the thermal column with other users, principally

the radiography group and the activation analysis group.

4.1.4. Source disk assembly . The detector capsule is now ready for

assembly with the source disks. Two pairs of source disks are presently in use (CV-3 through CV-6). Each source disk (15.9-mm dia. x 0.15-mm thick) is

made up of two 0.08-mm thick enriched uranium metal foils wrapped in 0.04-mm

thick aluminum foil. The total weight of a wrapped source disk is about

0.60 g and the total thickness is 0.38 mm. During irradiation the fission

neutron source strength of the two disks is about 4 x lO^^ n/s.

When not in use the source disks are kept at the far end of the lead

shield cart inside of the lead cup (LC). (To get LC in and out remove lead

bricks and short cover plate at front of cart.) Always check the radiation

level of the source disks before handling. Use the lead cup as a carrier when

moving source disks to the assembly area. Wear glasses for beta shielding .

4.1.5. Set up fission chamber monitor and electronics. Run "timer"

program on HP85 (see Appendix C). Check time marking instrument with

telephone time (844-2525) or Naval Observatory Time (653-0258). Notify

reactor control room. Raise the boral curtain while noting the time. Record

the time in notebook, irradiation log and worksheet (see appendix B), and

thermal column log sheet.

4.1.6. Leave sign with your name and phone numbers on boral curtain

control s

4.1.7. At the end of the irradiation (EOI), again check time. Notify

reactor control room. Lower the boral curtain while noting the time. Record

the EOI time in notebook, worksheet, and thermal column log.

4.1.8. Remove sensors from cadmium detector capsule using gloves and

package individually in marked plastic containers. Store in shielded card

file in building 235, room B123. Aluminum spacers and detector capsule should

be stored in pig either in room B123 or in thermal column area. Cadmium box

should be returned to its plexiglass holder and also stored as appropriate to

its residual radiation level.

4.2. FLUENCE AND FLUENCE RATE CALCULATIONS

4,2,1, Monitor Counting , The gamma radiation from the nickel (or

indium) foils is counted in the standard geometry with the intrinsic germanium

detector. The standard geometry is foil holder #3 mounted directly on the

detector with the foil located in the recessed hole, A pulser should be fed

into the preamplifier and the rate (usually 50 pps) adjusted to obtain sufficient counts in the pulser peak for the chosen counting interval. Detailed

instructions for the operation of the counting system are in appendix D, The

stored spectra are analyzed with the code STP2D8, (a FORTRAN code). A listing

of this code is in appendix E. The code includes a time history correction

and gives a final result as a counting rate at EOI. All results should be

corrected for deadtime and pileup by finding the live-time correction which

11
would restore the pulser results to the known, true pulser rate. The determined counting rate must be corrected for losses of events due to attenuation

of the outgoing gamma rays. The correction factor, C^, is equal to

(l-e"^^)/yt where y is the mass attenuation coefficient, for example, for the

810 keV line in nickel, y is 0.589 cm-i. The resulting correction for a .0254 cm nickel foil is 0.9925. This corrected result is then divided by the mass of the foil and entered as D in the worksheet,

4.2.2. Scattering corrections and gradient corrections . The FORTRAN

code "SCATFO" (see appendix F) is run on a Pro350 computer. Input data are taken from the foil assembly worksheet and the program gives u^^, the activation by scattered neutrons. The code also gives the field gradient,

(()( z)/(t)(0) . These parameters should be entered on the fluence calculation

worksheet. A value of D at the midplane of the foil holder can now be

calculated for each foil and an average value obtained for the set of foils.

4.2.3. Neutron Fluence Calculation . The code "SAT260" (see appendix F)

is run to obtain corrections for loss of activity during irradiation. If the

power level was approximately constant, a good correction can be obtained from

= (l-e"''^^)/xt where X is the decay constant and t is the length of the

irradiation. The activation decay rate factor, G, is equal to C^x. The

neutron midplane fluence is now calculated using equation 4 of the worksheet.

Table 6 of the worksheet can now be completed which gives the fluence and the

scattering correction for each sensor. Test reports should also be prepared

at this point for each sensor. A sample test report is in appendix H.

4.3. SUMMARY OF UNCERTAINTIES

The uncertainties of the certified fluence are given in Table 3. The

first part of the table gives uncertainties from the fluence transfer

procedure from the 252Qf field to the ^ssy field. The second part of the

table gives uncertainties for preparation of the fluence standards at the

cavity fission source. The resultant total uncertainty is for the free field

fluence. Additional uncertainties due to scattering corrections in the

fluence standard must also be added. This last term varies depending on the

activation reaction and the placement of the sensors in the detector capsule.

The table indicates which uncertainties are mostly systematic and which

are mostly random. In some cases, the uncertainty terms consist of both

random and systematic components. A more important distinction in the

uncertainty terms is whether the uncertainty could lead to an overall change

in scale of the NBS neutron fluence values. Two components could have this

effect: 1) the measured value of NBS I could change (thereby changing all source strength values, and 2) improved corrections for neutron scattering for

Pu(n,f) in the Cf and ^ssy fission fields.

12
TABLE 3. ASSIGNMENT OF UNCERTAINTIES ( ± 1 a)

Fluence Transfer Measurements from the ^^Kf Standard Fission

Neutron Field to the 235y Cavity Fission Source:

Percent

(1) Absolute neutron source strength of the 252cf source 0.9 S+R

(2) Distance measurement 0.6 R (3) 252cf to 235u spectrum-averaged cross section ratio

for Pu{n,f) monitor 0.1 S (4) Correction for neutron scattering for Pu(n,f) 0.9 S (5) Reproducibility of fluence transfer procedure (including

counting statistics, and electronic system stability

over period of measurement) 0.4 R

Subtotal (summed in quadrature) = 1.5

Preparation of Neutron Fluence Standards at the

235u Cavity Fission Source:

(1) Gradient corrections and positioning 0.4 S+R

(2) Corrections for neutron scattering for ii5in(n,n') 1.2 R (3) Reproducibility of fluence monitoring including random

summing corrections and uncertainties in activation

decay corrections 0.4 R

Subtotal (summed in quadrature)

TOTAL UNCERTAINTY = 2.0

S is for systematic uncertainty and R is for random uncertainty.

13
5. BENCHMARK REFERENCING

A benchmark neutron field is a wel 1 -character! zed neutron field which

will provide an accurate neutron fluence for validation or calibration of

experimental techniques and methods. For purposes of fluence transfer to a study field, the benchmark field should have a neutron energy spectrum that is

similar to the dosimetry environment to be monitored. Exposing identical sensors in both the benchmark field and the study field and hence determining

reaction rate ratios, eliminates a number of systematic errors (detector

efficiency, gamma attenuation, absolute cross section values, etc.) in the

determination of the unknown fluence.

Exposure of a dosimetry detector to a known neutron fluence, X25» ^ 235u fission spectrum establishes a calibration factor.

which may be used in conjunction with the cal cul ated fission-spectrum-averaged

cross section, a , for the detector reaction to obtain a consistent benchmark

A. calibration. The cross section, a , should be calculated with (1) the

A. fission spectrum shape associated with the a priori spectrum (e.g,, the source

spectrum if the a priori spectrum is from a neutron transport calculation);

and (2) the same energy dependent reaction cross section, R(E), used to

calculate and (commonly from the ENDF/B-V dosimetry file).

[If the absolute detection efficiency for the dosimetry measurement

method has been established (e.g., a good resolution fission chamber with

known deposit mass, or a calibrated gamma-ray counting system and known

reaction parameters, branching ratios, isotopic abundances, fission yields

etc.), then the "observed detector response" in Eq. (1) becomes the reaction

probability, and G is an observed cross section for the fission spectrum.]

The calibration factor may then be applied to the investigation of a study

spectrum.

By irradiating samples nearly identical to those irradiated in the ^ssy

field, the fluence of the study field, $ctu(jw» can be obtained as a simple

ratio of detector responses, D, and spectrum averaged cross sections, a, so that:

[observed detector response]

[xoc neutron fluence] (1)

D study _ ^studv study

(2)

D a

X X X

14
In practice, ^Q'^ ^'y is simply the time history corrected counting rate

ratio of the sample irradiated in the unknown field to that irradiated in the

235u field. Uncertainties due to isotopic abundances, nuclear parameters,

detector efficiencies, etc., disappear. The cross section ratios must be

calculated, but the absolute value becomes unimportant. The value of <i> is

that given in the test report and one then solves equation (2) for •J's-tudy

6. SAFETY

NRC and NBS regulations require that all personnel working at the NBS

reactor undergo a radiation safety and awareness training course. This course acquaints the worker with the applicable radiation regulations as well as the

hazards of radiation. Further training of any workers using the thermal

column facilities is also mandatory. Very high radiation doses can be

obtained through carelessness in the vicinity of the thermal column. A

specific hazard for the cavity fission source irradiations is the fissionproduct gamma activity from the ^ssy disks after they have been irradiated.

Radiation doses can be reduced by: (1) waiting as long as possible before

handling the disks, (2) having a well thought out plan of action for handling

the disks, (3) working rapidly, and (4) using shielding whenever possible.

Since the latter two often conflict, personal judgment based on experience

must be used. Item 2 can be facilitated by performing "dry runs" with "cold"

or dummy sources.

The irradiated samples will be radioactive, but at a very low level

(usually a few mi crocuri es) . There is therefore little hazard from the

samples, but all radioactive shipping regulations must still be followed. The

samples should be stored in a lead shielded storage area while awaiting

analysis or shipping.

15
REFERENCES

[1] DeJuren, J. A., Padgett, D. W., and Curtiss, L. F., J. Res. NBS 40, 63

(1955)

[2] "NBS Calibration Services, Users Guide 1986-1988." NBS SP250, U.S.

Government Printing Office, Washington, D.C. (1986).

[3] McGarry, E. D., Lamaze, G. P., Eisenhauer, C. M., Gilliam, D. M., and

Schima, F. J., "NBS ISNF and Cavity Fission U-235 Standard Neutron

Fields," Proc. Fourth ASTM-EURATOM Symposium on Reactor Dosimetry,

Washington, D.C, NUREG/CP-0029 , Vol. 2, 597 (March, 1982).

[4] McGarry, E. 0., Eisenhauer, C. M., Gilliam, 0. M., Grundl , J., and

Lamaze, G. P., "The U.S. U-235 Fission Spectrum Standard Neutron Field

Revisited," Proceedings Fifth ASTM-EURATOM Symposium on Reactor

Dosimetry, Geesthacht, Germany (September, 1984).

[5] Lamaze, G. P., McGarry, E.D., and Schima, F. J., "Integral Reaction

Rate Measurements in 252cf and 235u Fission Spectra," Proc. Int. Conf.

on Nuclear Data for Science and Technology, Antwerp, Belgium

(September, 1982).

[6] Eisenhauer, C. M. and Grundl, J. A., "Neutron Transport Calculations for

the Intermediate-Energy Standard Neutron Field (ISNF) at the National

Bureau of Standards," Proceedings International Symposium on Neutron

Standards and Applications, NBS Special Publication 493, U. S. Dept. of

Commerce, Washington, D.C. (March, 1977).

[7] Fabry, A. and Jenkins, J. D., "Wall Return Neutron Fluxes for High and

Intermediate-Energy Cavity Neutron Sources," Trans. Am. Nucl . Soc.

15, 2 (1972).

[8] Fabry, A., DeLeeuw, G., and DeLeeuw, S., CEN/SCK, Mol , Belgium, "The

Secondary Intermediate-Energy Standard Neutron Field at the Mol-zz

Facility," Nucl. Tech. 25, 349 (February, 1975).

[9] Fabry, A., Minsart, G., Cops, F., and DeLeeuw, S., "The Mol Cavity

Fission Spectrum Standard Neutron Field and Its Applications," Proc.

Fourth ASTM-EURATOM Symposium on Reactor Dosimetry, Washington, D.C,

NUREG/CP-0029, Vol. 2, 665 (March, 1982).

[10] Grundl, J., "Cavity Fission Spectra," Trans. Am. Nucl. Soc. 5^, 381

(1962).

[11] DeJuren, J. A. and Chin, J., J. Res. NBS 40, 311 (1955).

[12] Noyce, R. H., Mosburg, E. R., Jr., Garfinkel, J. B., and Caswell, R. S.,

"Absolute Calibration of the National Bureau of Standards Photoneutron

Source-Ill, Absorption in a Heavy Water Solution of Manganese Sulfate,"

Reactor Sci . Technol . (J. Nucl. Energy, Part A/B) \7_, 313 (1963).

[13] Spiegel, V., private communication.

16
[14] McGarry, E. D. and Boswel 1 , E. W., "Neutron Source Strength by the

Manganese Sulfate Bath Method," NBSIR, to be published.

[15] Eisenhauer, C. M., private communication.

[16] Madland, D. G., and Nix, J. R., "New Calculation of Prompt Fission

Neutron Spectra and Average Prompt Neutron Multiplicities," Nucl . Sci

Eng. 81, 213-271 (1982)

[17] Grundl , J. and Eisenhauer, C, "Fission Rate Measurements for Materials

Neutron Dosimetry in Reactor Environments," Proceedings First

ASTM-EURATOM Symposium on Reactor Dosimetry, EUR5667 e/f, Commission of

the European Committees, Petten (September, 1975).

[18] Grundl, J. A. and Eisenhauer, C, M., "Fission Spectrum Neutrons for Cross

Section Validation and Neutron Flux Transfer," Proceedings of a Conference on Nuclear Cross Sections and Technology, NBS Special Publication 425, U.S. Dept. of Commerce, Washington, D.C. (March, 1975).

[19] Wiedling, T., Proc. Symp. on Neutron Standards and Flux Normalization,

ANL (1970); see also Smith, A. B., Proc. Consultants Meeting on Prompt

Fission Neutron Spectra, Vienna (1971); Koster, A., Ibid., Vienna

(1971).

[20] Grundl, J., and Eisenhauer, C, "Compendium of Benchmark Neutron Fields

for Reactor Dosimetry," NBSIR 85-3151 (January, 1986).

[21] Magurno, B. A., Kinsey, R. R., and Schef fel , F. M., Guidebook for the

ENDF/B-V Nuclear Data Files, BNL-NCS-31451 (1982).

[22] Lamaze, G. P., Gilliam, D.M., and McGarry, E.D., "Neutron Fluence and

Cross Section Measurents for Fast Neutron Dosimetry," Proc. Conf.

Nuclear Methods and Environmental and Energy Research, Puerto Rico,

(April, 1984).

17

Appendix A

DETTECTOR ASSEMBLY WORKSHEET FOR CAVITY FISSION SOURCE

Assembly Date: 1/17/86 irradiation 8601 eOI: 21:37:00 Jan 20, 1986

501:11:01:00 EST, Jan 18, 1986

^Assembly Piece Depth Measurement S (mils) z (239-S)

Ident" thk. Diece sum micro" assigned (D- ) if icat ion meter = 324H19

239 Z = b -iZO

S pS C G IT 70

Au-l 5 167 72(Au-l) -48

Q n a r" o 1" 1 0

Ni-3 11 -L'-t 1 liRKj 91(Ni-3) -29

143 142.5

Spacer 11

Ni-4 11 126 113(Ni-4) -7

121

Spacer 11

Ni-5 11 105 134(ni-5) +15

99 101

Spacer 11

Ni-6 11 83 159(Ni-6) +^0

77 80

Spacer 15

S-10 6 59 180(Sc) +61

Au-2 5 53 186 (Au) +67

51 55

Spacer 60

E=249

Assembled Cd. Capjjule: 324 = f

Cd. Capsule Holder: 367
Pieces are assembled with l.d. number up

(2) Error does not Include uncertainty in midplane position. Position

error for Indium monitor is relative to adjacent foil.

f - dimension of assembled cadmium capsule (0.326** empty)

D - source disk separation: (f 119)

Z - S (102 - D/2)

- detector position measured from midplane

Foil Mass (g)

Ni-3 0. 28609

Ni-4 0, 28775

Ni-5 0 28564

Ni-6 0 28721

Au-1 0 28298

Au-2 0 29142

Sc S-10 0 04824

A2
Appendix B

IRRADIATION LOG AND WORKSHEET FOR CALCULATING NEUTRON FLUENCE

Irradiation Identification: 8601

Materials Irradiated : Au , Ni, Sc

1 . Source Assembly i.d. R/hr § 1 Ocm

Top Source Disk:

Bottom Source Disk:

Source Separation Distance: 4^3

2. Irradiation Log:

SOI: 11:01:00 EST Jan 18, 1986 (DAY l8 )

EOI: 21:37:00 EST Jan 20, 1986 (DAY 20)

Duration 2 days. 10 hr, 36 min = 210960S

Fission chamber monitor: Not used

3. Nickel and Indium Fluence Monitor Response at Midplane

D(c/s,g)

D(c/s,g) ^ '^Activation by § EOI

i.d. z 'i>(z)/<|)(0) § EOI Scattered Flux (corre c ted)

Ni-3 -29 1.0178 5491.1 -0.007 5^33

Ni-4 -7 1.0010 5395.1 -0.007 5428

Ni-5 +15 1.0048 5368.0 -0.007 5380

Ni-6 +40 1.0339 5436.0 -0.007 5295

5385±33(0.655)

B-l
H. Activation decay rate factors:

T= 2.10960S ^^Co

A= 1.133^x10-'

satfac 2.363x10-2

G= satfac/T 1 .1200x10-^

C= G/A = .9881

Note: For a constant fluence rate irradiation, satfac = (

1 -e

5. Fission Neutron Fluence at Midplane (z = 0)0

i^^Au

2.9762x10-^

4.663x10-1

2.210x10-^

.7426

''^Sc

9.5700x10-^

1 .999x10-2

9.474x10-^

.9900

XT.

' = K „ o cf o(x,.

)

[d/g]

counter response of fluence monitors

corrected for saturation (SATFAC code)

calculated cross section ratio for X to x^^

fluence transfer reaction (before 9-§3: In(n,n);

after 9-83:this factor is included in K ^) cf

calibration factor (i.e. composite of detector

efficiency parameters) ( For GEM detector, single shelf, zero thickness nickel foil, K ^ = 1.711 x 10 ) cf

<!> =1.711 xlO^

o 5385

,= 8.226x10^

1 .12x10-^

6. Fission Neutron Fluence at Neutron-Fluence-Standards Position

i. d. Z mils

$(z)/$(0) free-field fluence

§ foil pos.

y activity by

sci?tered flux(^)

Ni-3 -29 1 .01 78 8.373x10 1

-.7

Ni-4 -7 1 .0010 8.235x10 1 = -.7

Ni-5 + 15 1 .0048 8.266x10 ^^ -.7

Ni-6 + 40 1 .0339 8.506x10^ ^ -.7

Au-1 -48 1 .0488 8.628x10^ ^ + .4

Au-2 +67 1 .0952 9.010x10^ ^ + .5

S-10 +61 1 .0789 8.876x10 15 + .6

B-2
NOTES FOR FLUENCE CALCULATION

(a) Disk geometry gradient (empirical fit of calculation and measurement

)

4)(z) = neutron fluence rate per source neutron at z = distance from midplane

between coaxial disks a distance D apart.

= 0.0470 d"^ '^^

(1 + 21 .2 z^)

^ = 2 [1 . 1/(21.2 z2)]-l

^ cp z

D = 0.0470 "^^

(b) Corrections for scattered neutrons

Monte Carlo calculations from LANL provide net detector activation

associated with collided neutrons:

B-3

Appendix C

THESE ARE THE INSTRUCTIONS FOR RUNNING THE ADCAM DIRECTLY FROM PRO/TOOL KIT

1 . LOAD ADCAM MENU BY TYPING @MENU OR RUN DIRECTLY BY TYPING @START AND SKIP

TO STEP 3.1

2. THE "A" MENU WILL BE DISPLAYED; A CARRIAGE RETURN WILL CHANGE TO THE

"B" MENU; ANOTHER ONE CHANGES YOU BACK

3. FROM THE "B" MENU, TYPE "3" OR "START"

3.1 ANSWER QUERIES ABOUT MCA#

4. WHEN IT ASKS FOR PARAMETERS, TYPE LT

5. THEN TYPE IN THE PRESET TIME IN SECONDS

6. WHEN IT ASKS IF YOU WANT TO TRANSFER PARAMETERS FROM ANOTHER FILE,

SELECT NO (DEFAULT IS NO)

7. WHEN IT ASKS IF YOU WANT TO ALTER SAMPLE DESCRIPTION, TYPE Y

8. ENTER TWO LINE DESCRIPTION (DETECTOR DESCRIPTION IS ALREADY ENTERED

IN ANOTHER PLACE)

9. MAKE SURE SAMPLE IS IN PLACE!

10. WHEN IT ASKS IF YOU WANT TO ALTER ANALYSIS PARAMETERS, TYPE N OR CR

(DEFAULT)

11. DATA TAKING HAS NOW BEGUN, TIME IS AUTOMATICALLY RECORDED AND DATA

TAKING WILL END AT PRESET

12. YOU MAY DO OTHER THINGS WITH COMPUTER AT THIS TIME, e.g., UPDATE

GEML0G.DOC; TYPE CONTROL Z TO RETURN TO TOOLKIT FROM ADCAM MENU MODE

13. TO SEE DATA, SELECT EMULATE FROM "A" MENU, OR TYPE ©EMULATE FROM

TOOLKIT. FOR ADCAM 0 ANSWER ALL QUERIES WITH CARRIAGE RETURNS.

14. TO STORE DATA, ENTER EMULATE MODE (STEP 13)

14.1 HIT MENU KEY

14.2 SELECT SETUP MCA PARAMETERS

14.3 SET PRIMARY SOURCE TO MCB

14.4 SELECT ADCAM # AND SEGMENT # (1)

14.5 WHEN DONE, HIT EXIT, SELECT SAVE PRIMARY SPECTRUM, GIVE NAME, DO

15. WHEN DONE, HIT EXIT TO RETURN TO STEP ONE FOR NEXT SPECTRUM

16. PLEASE UPDATE GEML0G.DOC WITH NAMES OF STORED SPECTRA. PUT NAME OF

FLOPPY IN COMMENT SECTION IF NOT ON HARD DISC.

C-1

Appendix D

C STP2D8.FTN

C MODIFIED 10/1/84 TO MAKE VARIABLE Y 'VIRTUAL' FOR 8000 CHANNELS

C FORMERLY STPBG.FOR FROM PDP-11 RT-1 1 OPERATING SYS.

C MODIFIED 1984 FOR ADCAM FILES, I.E., .SPC FILES

C MODIFIED 9/25/82 BY MCGARRY (HAD TO CHANGE READ STATE IN READFI

)

C BACKGROUND IS COMPUTED FROM AVERAGES OVER REGIONS TO THE LEFT AND

C TO THE RIGHT OF THE PEAK REGION, RESPECTIVELY. THE BACKGROUND

C REGIONS NEED NOT BUTT UP AGAINST THE PEAK AREA.

C THIS PROGRAM WILL DO AN ENERGY CALIBRATION, LOCATE PEAKS,

C CALCULATE AREAS, WIDTHS, PERCENT ERRORS AND COUNTS PER SECOND.

C THE DATA OF THE SPECTRA MUST BE STORED ON A FILE WITH ANY STANDARD

C ORTEC .SPC FILE SPECIFICATION. PEAK AND END OF IRRADC lATION PARAMETERS ARE ENTERED IN SUBROUTINE PARAMS AND ARE READ

C FROM FILE LIMITS.DOC OR FROM THE KEYBOARD OR JUST NOT USED. WHEN

C THEY ARE NOT USED, THE PROGRAM WILL FIND THE PEAK PARAMETERS AND

C OUTPUT THEM ON FILE LIMITS. PKS. IF THEY ARE INPUT FROM THE KEYBOARD

C THEY WILL ALSO BE OUTPUT TO LIMITS. TEM. THEY SHOULD BE CONCATENATED

C TO LIMITS.DOC, IF ONE WISHES TO SAVE THEM. THE PROGRAM WILL STORE

C THE INTERCEPT, SLOPE AND SECOND ORDER CONSTANT FOR THE ENERGY CALIC BRATION CURVE ON KEV.DOC FROM THE ENERGY CALIBRATION ROUTINE OR USE

C AN EXISTING FILE WITH THAT DATA. THE RESULTS ARE WRITTEN ON THE FILEC NAME. RES AS WELL AS ECHOED TO THE TERMINAL IN SUBROUTINE EPEAK.

C

CC THE MAIN PROGRAM CALLS SUBROUTINES PARAMS, GETFIL,

C READFI, CAL, LOCATE, AND EPEAK IN THAT ORDER.

C

C

DIMENSION IWL(40) ,IWH(40) ,PK(40)

,

&IC(40) ,ID(40) ,NL(40) ,NR(40)

CHARACTER IFILI ( 20 ) , IFILO ( 20 ), REP

LOGICAL TEM

DIMENSION IWORD(34)

CHARACTER*128 TITL1 ,TITL2 ,ADATA

CHARACTER*5 AMONTH

CHARACTER*! ADUM

DOUBLE PRECISION XME(4)

VIRTUAL Y(8192)

L0GICAL*1 PRMS

COMMON/ADCAM/I ADCAM

COMMON /NCHAN/ ICHAN

COMMON /PKLIM/ NUM , IWL , IWH , IC , I D, NL , NR , PK , IBW , SIG

COMMON /FILNMS/ IFILI, IFILO

COMMON/IPUT/XME

COMMON/LAMBD/IYRO , IDAYO ,IHRO ,MINITO ,SECO , ALAM, CASRAT , &IYR,IDAY,IHR,MINIT,SEC,PRMS

COMMON /SPEC/TIME, CLTM,PLSRTE

2 F0RMAT(A1

)

3 FORMAT (14)

D-1
4 FORMAT (II

)

5 FORMAT C.P') ! THIS IS USED BY PGM EDPAGE

TYPE 8

8 FORMAT(/' LOCK SHIFT KEY'/)

CALL PARAMS(TEM)

10 CALL GETFIL(5, IFILI, IFILO)

C FORMERLY SUBROUTINE READFI

C

C READ AND PRINT TITLE AND COLUMN HEADS AND INPUT SPECTRUM DATA

C FROM THE INPUT FILE. IT IS ONLY CALLED BY THE MAIN PROGRAM.

C

READ(3' 1 ) (IWORD(I) ,1=1 ,34)

READ(3'IWORD(5))ADATA

PRINT *,ADATA

READ( ADATA ,7 ) ADATA( : 1 6 ), NDAY , AMONTH ,NYR , IHR , ADUM, MINIT , ADUM,

1NSEC,TIME,CLTM

7 F0RMAT(A16,I2,A5,I2,I5,A1 ,I2,A1 ,I2,2F11 .0)

PRINT 7 , ADATA ( : 1 6 )

, ND AY , AMONTH , NYR , IHR , ADUM , MINIT , ADUM , NSEC

,

1TIME,CLTM

WRITE (4 ,7) ADATA ( : 1 6 ) ,ND AY , AMONTH ,NYR , IHR , ADUM, MINIT , ADUM , NSEC

,

1TIME,CLTM

SEC=NSEC

IYR=1900+NYR

NLP=IM0D(NYR,4)

IF(AMONTH.EQ. ' -JAN- ' ) IM0N=1

IF(AMONTH.EQ. '-FEB-' )IM0N=2

IF ( AMONTH. EQ. ' -MAR- ' ) IM0N=3

IF(AMONTH.EQ. ' -APR- ' ) IM0N=4

IF( AMONTH. EQ. ' -MAY- ' ) IM0N=5

IF(AMONTH.EQ. ' - JUN- ' ) IM0N=6

IF(AMONTH.EQ. ' -JUL- ' ) IM0N=7

IF(AMONTH.EQ. '-AUG-' )IM0N=8

IF(AMONTH.EQ. ' -SEP- ' ) IM0N=9

IFCAMONTH.EQ. '-OCT-' )IM0N=10

IF(AMONTH.EQ. '-NOV-' )IM0N=1

IF( AMONTH. EQ. ' -DEC- ' ) IM0N=1

IDAY=(IM0N-1 )*30+NDAY

IF(IMON.EQ.2.0R.IMON.EQ.6.0R.IMON.EQ.7)IDAY=IDAY+1

IF(IM0N.EQ.3)IDAY=IDAY-1

IF(IM0N.EQ.8)IDAY=IDAY+2

IF(IM0N.EQ.9.0R.IM0N.EQ. 10)IDAY=IDAY+3

IFdMON.EQ. 1 1 .OR.IMON.EQ. 12)IDAY=IDAY + 4 C IF LEAP YEAR(NLP=0)ADD A DAY AFTER FEBRUARY

IF(NLP.EQ.0.AND.IMON.GE.3)IDAY=IDAY+1

READ(3'IWORD(6))TITL1

READ(3'IWORD(7))TITL2

PRINT *,TITL1 ,TITL2

WRITE(4,ilO)TITL1 ,TITL2

40 F0RMAT(4(1 X,A64/)

)

INC=0

DO 20 IN=IW0RD(31 ) ,IW0RD(31 )+255

READ(3'IN)(Y(I),I=INC+1 ,INC+32)

20 INC=INC+32

ICHAN=INC

D-2
C FORMERLY END ! OF SUBROUTINE READFI

D WRITE(4,5)

21 TYPE 200

200 FORMATC ENTER ADCAM NUMBER(0 OR 1)')

ACCEPT 4,IADCAM

IF (lADCAM.NE.O .AND. lADCAM. NE. 1 ) GO TO 21 TYPE 9

9 FORMAT(/' THE ENERGY CALIBRATION DATA SHOULD BE ON KEV.DOC'//

&'$D0 YOU WANT A NEW ENERGY CALIBRATION ? : ')

ACCEPT 2, REP

IF(REP.EQ. ' Y' ) CALL CAL(Y)

TYPE 13

13 FORMATC DO YOU WANT TO LOCATE PEAKS, USING THE ENERGY ' &//'$CALIBRATION FROM KEV.DOC ? ')

ACCEPT 2, REP

IF(REP.EQ. ' Y' ) CALL LOCATE(Y)

TYPE 12

12 FORMAT ('$D0 YOU WANT TO ANALYSE THE SPECTRUM : ')

ACCEPT 2, REP

IF(REP.EQ. ' Y' ) CALL EPEAK(Y)

TYPE 15

15 FORMATC DO YOU WANT TO INPUT ANOTHER FILE USING THE SAME',

&/'$ PARAMETERS AND INPUT FORMAT ? ')

CL0SE(3)

CL0SE(4)

ACCEPT 2, REP

IF(REP.EQ. 'Y' ) GO TO 10

TYPE 25, IFILO

25 F0RMAT(7X'THE RESULTS ARE ON FILE ' ,20A1

)

IF (TEM) GO TO 27

OPEN (UNIT=2,FILE=' LIMITS. PKS' ,STATUS=' OLD '

)

C THE PREVIOUS LINE MODIFIED FROM RT-1 1 VERSION WHICH USED CALL ASSIGN

DO 65 1=1 ,NUM

65 WRITE(2,105)IWL(I) ,IC(I) ,NL(I) ,ID(I) ,IWH(I) ,NR(I) ,PK(I)

105 FORMAT(6l8,F8.2)

CL0SE(2)

TYPE 28

28 FORMATC THE PEAK LIMITS HAVE BEEN WRITTEN ON LIMITS . PKS

)

GO TO 30

27 TYPE 26

26 F0RMAT(7X'D0 NOT FORGET FILE LIMITS. TEM ')

30 STOP

END ! OF MAIN PROGRAM

c

C

SUBROUTINE PARAMS(TEM)

C

C THIS SUBROUTINE PROVIDES THE INPUT FROM THE KEYBOARD OR FROM FILE

C LIMITS.DOC FOR READING THE VARIOUS PARAMETERS FOR THE PROGRAM.

C THOSE SUPPLIED FROM THE KEYBOARD WILL BE WRITTEN TO FILE LIMITS. TEM

C SO THAT THEY CAN BE CONCATENATED TO LIMITS.DOC LATER.

C IF ONE IS NOT INTERESTED IN PARAMETERS THE PROGRAM WILL

C SUPPLY A SET THAT WILL GIVE AN OUTPUT OF PEAK SUMS ONLY.

D-3
C IN THE CASE OF A DECAY LIKE BA , WHICH DECAYS TO LA, ONE MUST

C INPUT BOTH OF THE HALF LIVES, THE LONGEST FIRST (BA). IT CALCULATES

C THE CASCADING RATIO OF BA ACTIVITY TO LA ACTIVITY AT THE END OF

C IRRADIATION, (TA-TB)/TA.

C

DIMENSION IPK(40) ,IWL(40) , IWH(40) ,PK{HO)

,

&IC(MO) ,ID(40) ,NL(40) ,NR(ilO)

DOUBLE PRECISION KEY, TOKEN

CHARACTER REPL

LOGICAL*! PRMS,TEM

COMMON/LAMBD/IYRO , IDAYO , IHRO , MINITO ,SECO , ALAM, CASRAT

,

&IYR,IDAY,IHR,MINIT,SEC,PRMS

COMMON /FILNMS/ IFILI.IFILO

COMMON /NCHAN/ ICHAN

COMMON /PKLIM/ NUM , IWL , IWH , IC , I D, NL , NR , PK , IBW , SIG

DATA NUM, IYRO,IYR, IDAYO, IDAY, IHRO, IHR, MINITO, MINIT /9*0/,

&TA /1.0 E13 /, TB,SECO,SEC /3*0 . 0/ , PRMS/ . FALSE . / , I BW/3/

,

&SIG /40./

TEM=. FALSE.

TYPE 1

1 FORMAT (

' $D0 YOU WANT TO USE INTEGRATION PARAMETERS ? '

)

2 FORMATCl A1

)

ACCEPT 2, REPL

IF(REPL.NE. ' Y' ) GO TO 120

PRMS = .TRUE.

TYPE 8

8 FORMAT ('$D0 YOU WISH TO READ THEM FROM LIMITS.DOC ? ')

ACCEPT 2, REPL

- IF (REPL.EQ. 'Y' ) GO TO 100

TYPE 9

9 FORMATC THESE PARAMETERS WILL BE WRITTEN ON FILE LIMITS. TEM',

&/' SO THAT YOU CAN LATER ADD THEM TO FILE LIMITS.DOC)

OPEN (UNIT=2,FILE=' LIMITS. TEM' ,STATUS='NEW'

)

TYPE 101

ACCEPT 102, KEY

WRITE (2, 102) KEY

WRITE (5, 122) KEY

TEM = .TRUE.

TYPE 3

3 FORMATC GIVE YEAR , DAY , HR , MIN , SEC FOR THE END OF IRRADIATION ' &/' IN 4110, 1F10.0 FORMAT, SEPARATED BY COMMAS V)

ACCEPT 4 , lYRO , I DAYO , IHRO , MINITO ,SECO

TYPE 4, lYRO, IDAYO, IHRO, MINITO, SECO

4 FORMAT(4I10,F10.0)

TYPE 5

5 FORMAT(/' GIVE THE HALF LIFE OF THE SINGLE DECAY OR THE TWO HALF'

&/' LIVES FOR A CASCADING DECAY LIKE BA-LA , LONGEST FIRST AND IN'

&/' F OR El 4.0 FORMAT, SEPARATED BY A COMMA, IF TWO '/)

ACCEPT 6, TA, TB

TYPE 206, TA, TB

6 FORMAT (2F1 4.0)

206 F0RMAT(1X,2E14.6)

TYPE 7 7 FORMAT(/'$UNITS OF TIME FOR THE FIRST (Y, D, H, M, OR S) ? ')

D-4
ACCEPT 2,REPL

IF(REPL.EQ. 'Y' ) GO TO 20

IF(REPL.EQ. 'D' ) GO TO 21 IF(REPL.EQ. 'H' ) GO TO 22

IF(REPL.EQ. 'M' ) GO TO 23

IF(REPL.EQ. 'S' ) GO TO 26

20 TA=TA*365.25

21 TA=TA*2i|,

22 TA=TA*60.

23 TA=TA*60.

26 IF(TB.LT.1 .OE-5) GO TO 50

27 TYPE 28

28 FORMAT (' $UNITS OF TIME FOR THE SECOND ( Y, D, H , M, OR S) ? ')

ACCEPT 2,REPL

IF(REPL.EQ. 'Y' ) GO TO 30

IF(REPL.EQ. 'D' ) GO TO 31 IF(REPL.EQ. 'H' ) GO TO 32

IF(REPL.EQ. 'M' ) GO TO 33

IF(REPL.EQ. 'S' ) GO TO 50

30 TB=TB*365.25

31 TB=TB*2i|.

32 TB=TB*60.

33 TB=TB*60.

50 TYPE 51

51 FORMAT (' HOW MANY PEAKS DO YOU WISH TO USE (12) ')

ACCEPT 53,NUM

TYPE 52

52 FORMAT(/' $PEAK SIGNIFICANCE : ')

ACCEPT 54,SIG ! USED AFTER ST 70 IN SUB CHPEAK

5M FORMAT (F6.0)

53 FORMAT (16)

55 FORMAT ( 5 X,' ENTER THE LEFT BKG. CHANNEL, PEAK CHANNEL AND',/,

&' THE NUMBER OF BKG. CHANNELS TO BE AVERAGED ON THE LEFT. ENTER',/

&' UPPER PEAK CHANNEL ,THE BKG. CHANNEL, AND NUMBER OF CHANNELS ',/

&' TO BE AVERAGED ON THE RIGHT. FOLLOW THIS WITH THE PEAK POSI-',/

&' TION. ALL ENTRIES IN 6l , IF FORMAT ON ONE LINE WITH A COMMA',/

&' BETWEEN EACH. '

)

TYPE 55

DO 63 1=1 ,NUM

TYPE 59,1

63 ACCEPT 61 ,IWL(I) ,IC(I) ,NL(I) ,ID(I) ,IWH(I) ,NR(I) ,PK(I)

59 F0RMAT(7X,' PEAK NO. ',I4)

61 FORMAT (6110, FU.O)

WRITE ( 2 , 1 OM ) NUM , I YRO , I DAYO , IHRO , MINITO , SECO , SIG , TA , TB

DO 65 1=1 ,NUM

65 WRITE(2,105)IWL(I) ,IC(I) ,NL(I) ,ID(I) ,IWH(I) ,NR(I) ,PK(I)

CL0SE(2)

GO TO 120

100 OPEN (UNIT=2,FILE=' LIMITS. DOC ,STATUS= ' OLD '

)

TYPE 101

101 FORMAT ( '$WHAT IS THE ISOTOPE KEYWORD ? ')

ACCEPT 102, KEY

102 F0RMAT(A8) ! DOUBLE PRECISION = 8 CHARACTERS

WRITE(5,122) KEY

D-5
122 F0RMAT(1X,A8)

103 READ(2,102,END=1 10) TOKEN

IF(TOKEN.NE.KEY)GO TO 103

104 FORMAT(5I8,F8.0,F8.0 , /,2E16.6)

READ( 2 , 1 04 )NUM, lYRO , IDAYO , IHRO ,MINITO ,SECO ,SIG,TA ,TB

WRITE(5 ,104)NUM,IYR0,IDAY0,IHR0,MINIT0,SEC0,SIG,TA,TB

DO 106 1=1 ,NUM

READ(2,105)IWL(I) ,IC(I) ,NL(I) ,ID(I),IWH(I) ,NR(I) ,PK(I)

106 WRITE(5 ,105)IWL(I) ,IC(I) ,NL(I) ,ID(I) ,IWH(I) ,NR(I) ,PK(I)

105 F0RMAT(6l8,F8.2)

CL0SE(2)

GO TO 120

1 10 TYPE 112

112 FORMAT (' KEY WAS NOT FOUND IN LIMITS.DOC FILE ')

120 ALAM=AL0G(2. ) /TA

CASRAT=(TA-TB)/TA

RETURN ! TO MAIN PROGRAM

END ! OF SUBROUTINE PARAMS

C

c

c

c

c SUBROUTINE GETFIL ( LUN , AREAI , AREAO

)

C

C THIS SUBROUTINE IS CALLED FROM MAIN IN ORDER TO ASSIGN

C THE INPUT AND OUTPUT FILES TO LOGICAL UNITS

C

CHARACTER*20 AREAI, AREAO

TYPE 1

1 FORMAT (' $ENTER INPUT FILENAME ')

READ(LUN,9) IQI, AREAI(:IQI)

IQ=IQI

K=0

DO 10 I = 1 ,IQ

10 IF (AREAI(I:I) .EQ. '.') K = I ! DECIMAL POINT

IF ( K. EQ.O) AREAI (IQ+1 :IQ+4)=' .SPC

U IF(K .NE. 0) IQ=K-1

AREA0( :IQ) = AREAI( :IQ)

AREA0(IQ+1 : IQ + 4) = ' .RES'

9 FORMAT (Q, A20)

WRITE (LUN, 8) AREAI ( : IQ+4 ), AREAO ( :IQ+4)

8 F0RMAT(2X,A20,3X,A20)

0PEN(UNIT=3,FILE=AREAI( : IQ+4 )

, STATUS= ' OLD ' , ACCESS= ' DIRECT'

,

1 REC0RDSIZE=32)

0PEN(UNIT=4,FILE=AREA0( : IQ+4 )

, STATUS= ' NEW '

)

RETURN ! TO MAIN PROGRAM

END ! OF SUBROUTINE GETFIL

D-6
c SUBROUTINE CAL(Y)

C

C SUBROUTINE GAL USES SUBROUTINE ELOC TO FIND PEAKS

C IT IS ONLY CALLED BY THE MAIN PROGRAM

C

1 1

12

10

20

30

40

45

DIMENSION X(40),E(40)

VIRTUAL Y(8192)

COMMON /PEAK/X,E,N

COMMON/ ADC AM/I ADC AM

DATA SX , SX2 , SX3 , SX4 , SY , SXY , SX2 Y/7*0 . 0/

GALL ELOC(Y)

TYPE 2

FORMAT (' PEAK CHANNEL ENERGY'/)

DO 1 1 1 = 1 ,N

TYPE 12,I,X(I),E(I)

CALCULATE PEAK ENERGY E = A + BX + CX**2

2)

X'S

SUM X**2

X**3

SUM X**4

SUM

SUM

SUM

Y'S

X*Y

Y*X**2

F0RMAT(I4,5X,F9.2 ,2X,F9

IF(N.EQ. 1 ) GO TO 20

DO 10 1=1 ,N

SX=SX+X(I) ! SUM

X2=X(I)*X(I)

SX2=SX2+X2

SX3=SX3+X2*X(I)

SX4=SX4+X2*X2

SY=SY+E(I)

SXY=SXY+X(I)*E(I)

SX2Y=SX2Y+X2*E(I)

CONTINUE

T=N

CKSX3=T*SX3-SX*SX2

CKSX4=T*SX4-SX2*SX2

CKSX2=T*SX2-SX*SX

CKSXY=T*SXY-SX*SY

CKSX2Y=T*SX2Y-SX2*SY

CI =CKSX2*CKSX2Y-CKSX3*CKSXY

C2=CKSX2*CKSX4-CKSX3*CKSX3

C=C1 /C2

B1=CKSXY-C*CKSX3

B=B1 /GKSX2

A1=SY-C*SX2-B*SX

A=A1/T

GO TO 30

A=0.0

0=0.0

B=E(1 )/X(1 ) TYPE 40

FORMAT(/' DATA FILE KEV.DOC

TYPE 45,A,B,C

FORMAT (' A, B,

THIS IS A CLOSED SOLUTION TO EQ (8-4)

OF BEVINGTONS'S DATA REDUCTION AND ERROR

ANALYSIS FOR THE PHYSICAL SCIENCES

/)

C = ' ,1P3E14.7)

IF(IADCAM.GT.0)GO TO 200

D-7
0PEN(UNIT=2,FILE='KEV0.D0C' ,STATUS= ' OLD '

)

GO TO 201

200 0PEN(UNIT=2,FILE='KEV1 . DOC ,STATUS= ' OLD '

)

201 WRITE(2,50)A,B,C

CL0SE(2)

50 FORMAT(3E17.10)

TYPE 55

55 F0RMAT(/' FILE KEV UPDATED. '/)

RETURN ! TO THE MAIN PROGRAM

END ! OF SUBROUTINE GAL

C

SUBROUTINE ELOC(Y)

C IT RETURNS PEAK LOCATIONS WITH THE ENERGY SUPPLIED BY THE

C TERMINAL FOR THE ENERGY CALIBRATION, WHICH IS DONE BY CAL.

C IT CALLS SUBROUTINE CHPEAK.

C

DIMENSION X(MO) ,E(40)

VIRTUAL Y(8192)

COMMON /NCHAN/ ICHAN

COMMON /PEAK/X,E,NUM

NUM=0

TYPE 1

1 FORMAT(/'$PEAK SIGNIFICANCE : ')

ACCEPT 2,SIG ! USED AFTER ST 70 IN SUB CHPEAK

2 FORMAT (F6.0)

TYPE 3 3 F0RMAT(//4X, 'PEAK LOCATOR'//' CHAN ENERGY '/)

1 = 1 10 CALL CHPEAK(CPK,IH,IL,I ,SIG, Y)

IF(CPK .LT. 0) GO TO 100

NUM=NUM+1

X(NUM)=CPK

TYPE 4,CPK

4 FORMATdH ,$F7.1 ,2X/)

ACCEPT 5,E(NUM)

5 FORMAT (F1 0.0)

IF(E(NUM) .LT.1 .0) NUM=NUM-1

100 IFd.LT. ICHAN) GO TO 10

RETURN ! TO SUBROUTINE CAL

END ! OF SUBROUTINE ELOC

C

c SUBROUTINE CHPEAK( CPK , IH , IL , I ,SIG, Y)

C

C THIS SUBROUTINE LOCATES THE PEAK FROM THE INPUT DATA.

C IT STARTS AT CHANNEL I AND STEPS FROM CHANNEL TO CHANNEL

C LOOKING FOR PEAKS. IT IS CALLED BY SUBROUTINES ELOC, EPEAK

C AND LOCATE. A PEAK MUST BE AT LEAST TWO CHANNELS WIDE IN WHICH

C CASE THE SECOND CHANNEL COUNT MUST BE LARGER THAN THE FIRST.

C THE ALTERNATIVE IS TO ENTER THE REGION OF INTEREST INTO THE

C LIMITS.DOC FILE, WHICH IS ACCESSED BY SUBROUTINE PARAMS, AND

D-8
C ANSWER NO TO THE GAINSHIFT REQUEST IN SUBROUTINE EPEAK. THEN

C THE PROGRAM WILL NOT ACCESS THIS SUBROUTINE.

C

VIRTUAL Y(8192)

COMMON /SPEC/ TIME.CLTM

ISTEP=1

I=I+ISTEP

IF (Y(I).LT.O. )Y(I)=-Y(I)

YL=Y(I-1 )+SQRT(Y(I))

YH=Y(I+1 )-SQRT(Y(I))

IF((YL.LT.Y(I)).AND.(Y(I).LT.YH)) GO TO 50

GO TO ICQ

50 IL=I

52 DIF=Y(I+1 )-Y(I)

IF(DIF.LT.O.O) GO TO 51

1=1 + 1 GO TO 52

51 YP=Y(I)

IK=I

53 1=1 + 1 YL=Y(I-1 )+SQRT(Y(I))

YH=Y(I+1 )-SQRT(Y(I))

IF((YL.GT.Y(I)).AND.(Y(I).GT.YH)) GO TO 53

54 IH=I

YL=Y(IL)

YH=Y(IH)

BK=(YL+YH)/2.0

PW=IH-IL

IF(PW.LE.3) GO TO 100

70 CNT=YP-BK

TOT=YP+BK

IF(TOT.LE.O. ) GO TO 100

IF(CNT.LT.SIG*SQRT(TOT) ) GO TO 100 ! SIG = PEAK SIGNIFICANCE

SXY=0. ! SELECTION CRITERION

SY=0.

IL0W=IK-2

IHI=IK+2

DO 75 K=ILOW,IHI

CK=K

SXY=SXY+CK*Y(K)*Y(K)

SY=SY+Y(K)*Y(K)

75 CONTINUE

CPK=SXY/SY

GO TO 101

100 CPK= -10.0

101 RETURN ! TO SUBROUTINE ELOC , LOCATE, OR EPEAK

END ! END OF SUBROUTINE CHPEAK

C SUBROUTINE LOCATE (Y)

C

C SUBROUTINE LOCATE DETERMINES PEAKS AND BOUNDARIES AND

C PASSES INFORMATION TO SUBROUTINE EPEAK VIA COMMON BLOCKS.

C IT WRITES ON THE OUTPUT FILE FILENAME . RES

D-9
c DIMENSION EGY(40) ,IPK(40) ,IWL(40) ,IWH(40) ,PK(40)

,

&IC(40) ,ID(MO) ,NL(MO) ,NR(40)

VIRTUAL Y(8192)

COMMON/ ADC AM/I ADC AM

COMMON /NCHAN/ ICHAN

COMMON /ENEG/ EGY

COMMON /PKLIM/ NUM , IWL , IWH , IC , ID , NL ,NR , PK , IBW, SIG

IFdADCAM. GT.OGO TO 200

OPEN(UNIT=2,FILE='KEV0.DOC' ,STATUS= ' OLD '

)

GO TO 201

200 0PEN(UNIT=2,FILE='KEV1 .DOC ,STATUS= ' OLD '

)

201 READ(2,1 )A,Z,C

1 F0RMAT(3E1 7. 10)

CLOSE (2)

NUM=0

TYPE 2

2 FORMAT(/' $PEAK SIGNIFICANCE : ')

ACCEPT 3, SIG ! USED AFTER STMT 70 IN SUB CHPEAK

3 FORMAT (F6.0)

WRITE(4 ,4)

WRITE(4 ,i1) 4 FORMAT( ' **************************************

WRITE(4 ,5 ) TYPE 5

5 F0RMAT(4X, ' PEAK LOCATOR'//' PEAK CHAN ENERGY FWIDTH'/)

1 = 1 10 CALL CHPEAK(CPK,IH,IL,I,SIG,Y)

IF(CPK .LT. 0) GO TO 100

EN=A+Z*CPK+C*CPK*CPK

NUM=NUM+1

EGY(NUM)=EN

IFW=IH-IL+1

IWL(NUM)=IL

IWH(NUM)=IH

IC(NUM)=IL

ID(NUM)=IH

NL(NUM)=IBW

NR(NUM)=IBW

PK(NUM)=CPK

WRITE(4 ,6 )NUM,CPK,EN ,IFW

TYPE 6,NUM,CPK,EN,IFW

6 F0RMAT(I5,F8.2,F9.2,3X,I5)

100 IFd.LT. ICHAN) GO TO 10

RETURN ! TO SUBROIUTINE EPEAK

END ! OF SUBROUTINE LOCATE

Q*****************************it X** *****************5f*^t**************

C

SUBROUTINE EPEAK(Y)

C

C SUBROUTINE EPEAK USES PEAKS AND BOUNDARIES FOUND BY

C SUBROUTINE LOCATE OR THOSE FROM SUBROUTINE PARAMS

C IT CREATES ALL THE OUTPUT RESULTS WHICH ARE WRITTEN

D-10
C TO FILENAME. RES. IT WILL ADJUST THE PEAK PARAMETERS WITH RESPECT

C TO THE PEAK FOR CHANGES IN GAIN OF THE PEAK. IT CALLS SUBROUTINE

C CHPEAK AND LOCATE. IT IS CALLED BY THE MAIN PROGRAM.

C

DIMENSION EGY(40) , I PK( MO ) , I WL( 40 ) , IWH ( 40 ) , PK( ^JO ) , &IC(40) ,ID(40) ,NL(40) ,NR(40)

DIMENSION BL0W(40) ,BHIGH(40) ,BKG(40) ,T0TAL(40) ,ER(40)

CHARACTER IFILI(20) , IFILO( 20 )

, IREP

LOGICAL*! GAIN, PRMS, PULSE

DOUBLE PRECISION XME(4)

VIRTUAL Y(8192)

COMMON/ ADC AM/ I ADC AM

COMMON/I PUT/XME

COMMON/LAMBD/IYRO , IDAYO , IHRO , MINITO ,SECO , ALAM , CASRAT

,

&IYR,IDAY,IHR,MINIT,SEC,PRMS

COMMON /ENEG/ EGY

COMMON /SPEC/ TIME,CLTM,PLSRTE

COMMON /ENFIT/ A,Z,C

COMMON /PKLIM/ NUM , IWL , IWH , IC , I D, NL , NR , PK , IBW , SIG

COMMON /FILNMS/ IFILI.IFILO

DATA GAIN/. FALSE./

IFdADCAM. GT.0)GO TO 200

OPEN ( UNIT=2, FI LE= 'KEVO. DOC ' ,STATUS= ' OLD '

)

GO TO 201

200 0PEN(UNIT = 2,FILE='KEV1 . DOC ,STATUS= ' OLD '

)

201 READ(2,1 )A,Z,C

1 FORMAT(3E17.10)

2 F0RMAT(A1 ) CL0SE(2)

TYPE 22

22 FORMAT ('$D0 YOU WANT TO CORRECT FOR GAIN SHIFT ? ')

ACCEPT 2, IREP

IFCIREP.EQ. ' Y' ) GAIN = .TRUE.

C

C OUTPUT RESULTS

C

TYPE 3 WRITE(4 ,3)

3 FORMAT(' **************************************

C WRITE(M,4)IFILI

TYPE 4,IFILI

4 FORMAT(/' DATA FILENAME: '21A1)

WRITE(4,5)TIME

TYPE 5, TIME

5 FORMAT(/' LIVE TIME ( SEC ) : '

,

F8 . 0 , /

)

21 CONTINUE

7 FORMAT (' PEAK CHANNELS ENERGY BKGND AREA % ERR

& CTS/SEC'/)

9 FORMAT (/)

CC DETERMINE AREAS

C

D-11
IDAYS = 0

SECS = 0.0

IF (.NOT. PRMS) GO TO 72

10 IF(IYRO.LE.IYR)GO TO 70

WRITE(5,12)IYR0,IYR

12 FORMATC IYR0,',I5,', CANNOT BE LARGER THEN IYR,',I5)

TYPE 18

18 FORMATC INPUT NEW lYRO ')

ACCEPT 19,IYR0

19 FORMAT (I6)

GO TO 10

70 IFdYRO.EQ. IYR)GO TO 71 IF(M0D(IYR0,4) .EQ.O)IDAYS =IDAYS+1

IYR0=IYR0+1

IDAYS=IDAYS+365

GO TO 70

71 IDAYO=IDAYO-IDAYS

FDAYS=IDAY-IDAYO

FIHRS=FDAYS*24.+(IHR-IHR0)

FMINIT=FIHRS*60.+(MINIT-MINIT0)

SECS=FMINIT*60 . + (SEC-SECO

)

72 TYPE 88, SECS

WRITE(i| ,88)SECS

WRITE(i|,7)

TYPE 7 88 FORMATC SECONDS AFTER END OF IRRADIATION = \E^^.6,/)

PULSE = .FALSE.

IF(PLSRTE.LT.2) GO TO 885

PULSE = .TRUE. ! PULSER PEEK IN SPECTRUM

NNUM=NUM+1

DO 870 I=NNUM,2,-1

EGY(I) = EGY(I-1 ) IPK(I) = IPK(I-1 ) IWL(I) = IWL(I-1 ) IWH(I) = IWH(I-1 ) PK(I)= PK(I-1 ) IC(I)= IC(I-I)

ID(I)= ID(I-1 ) NL(I)= NL(I-I)

870 NR(I)= NR(I-1 ) EGY( 1

)

=EGY(NNUM)

IPKd ) =IPK(NNUM)

IWL(1

)

=IWL(NNUM)

IWH(I) =IWH(NNUM)

PK(1 )= PK(NNUM)

ICC ) = IC(NNUM)

IDC ) = ID(NNUM)

NLC ) = NL(NNUM)

NR ( 1 ) = NR(NNUM)

GO TO 886

885 PLSRTE=1 ! USED IN ST 86

PLSRAT=1 ! USED IN ST 86

886 DO 40 1=1 ,NUM

IDIFF=0

D-12
IF(.NOT. PRMS)GO TO 15

IF (.NOT. GAIN) GO TO 15

IF((PK(1 ) .LT.IWL(I)).OR. (PK(I).GT.IWH(I)) )TYPE 16

16 FORMAT (' INPUT PEAK IS OUT OF BOUNDS ')

LL=IWL(I)-NL(I)

ILL=LL

LH=IWH(I)+NR(I)

17 CALL CHPEAK(CPK,IH,IL,ILL,SIG, Y)

IF(CPK.GT.5.0)GO TO 14

IF(ILL.LT.LH)GO TO 17

TYPE 13,LL,LH,SIG

13 FORMAT (' NO PEAK WAS FOUND BETWEEN CHANNEL 15 , ' AND ',15,

&/' WITH PEAK SIGNIFICANCE =',F5.1)

GO TO 40

14 CPKK=CPK+0.5

TPK=PK(I)+0.5

IDIFF = IFIX(CPKK)-IFIX(TPK)

PK(I)=PK(I)+IDIFF

IWH(I)=IWH(I)+IDIFF

IWL(I)=IWL(I)+IDIFF

IC(I)=IC(I)+IDIFF

ID(I)=ID(I)+IDIFF

IF ( IDIFF. NE.O) WRITE (5 ,8) IDIFF

8 FORMATC THE PROGRAM PARAMETERS HAVE BEEN SHIFTED', 15,

&/' CHANNELS DUE TO A CHANGE IN PEAK POSITION.'/)

15 T=0.0

IMINS=IC(I)

IPLUS=ID(I)

CC=FLOAT(IC(I))

DD=FLOAT(ID(I))

DO 41 J=IMINS,IPLUS

T=T+Y(J)

41 CONTINUE

C

C PEAK LOCATION

C

SXY=0.

SY=0.

DO 45 K=IMINS,IPLUS

CK=K

SXY=SXY+CK*Y(K)*Y(K)

45 SY=SY+Y(K)*Y(K)

CPK=SXY/SY

C

C PROGRAM USES NL AND NR CHANNELS ON EACH SIDE TO DETERMINE

C BACKGROUND AT PEAK BOUNDARIES. DEFAULT IS IBW.

C

42 IMINS=IWL(I)

IPLUS=IWH(I)

BL=0.0

DO 64 JK-1 ,NL(I)

64 BL=BL+Y(IMINS-JK+1

)

NTL=NL(I)

XL=FLOAT(IMINS)-( FLOAT (NTL)/2,

)

D-13
BL=BL/FLOAT(NTL)

BLOW(I)=BL

BH=0.0

DO 66 JK=1 ,NR(I)

66 BH=BH+Y(IPLUS+JK-1

)

NTR=NR(I)

XH=FLOAT ( I PLUS ) + ( FLO AT ( NTR ) /2 . 0

)

BH=BH/FLOAT(NTR)

BHIGH(I)=BH

C

C BACKGROUND EQUATION

C

B=BL*(CPK-CC+0.5)+BH*(DD-CPK+0.5)

YAA=T-B

BKG(I)=B

TOTAL(I)=T

C

C PEAK ENERGY

C

EGY( I )=A+Z*CPK+C*CPK*CPK

C

C PERCENT ERROR = SQRT (T+ ( ( U-L) ^2) * (BL+BH ) ) /N

C

C0NST=(DD-CC)/2.

SUM=BL+BH

ER(I)=SQRT(T+CONST*SUM)/YAA

ER(I)=ER(I)*100.

BINT=YAA*CLTM/TIME

RATEUN=BINT/CLTM IDEADTIME CORRECTED RATE

C

C RATESC IS RATE CORRECTED FOR WITHIN COUNTING TIME DECAY

C FOR VERY SMALL CORRECTIONS, THIS CORRECTION IS NOT MADE

C

IF(.NOT. PULSE)GO TO 74

RATESC=BINT/CLTM

RATETO=RATESC

PLSRAT=RATESC ! ANALYZER PULSER COUNT

PULSE = .FALSE. ! (PLSRTE/PLSRAT) HAS BEEN DETERMINED FOR ST 86

CORR=PLSRTE/PLSRAT

GO TO 90

74 C0RR=1

IF(CLTM*ALAM .LT. .001)00 TO 87

RATESC=BINT*ALAM/(1 -EXP(-ALAM*CLTM)

)

GO TO 86

87 RATESC=BINT/CLTM

86 RATETO=CASRAT*RATESC*CORR*EXP(ALAM*SECS)

90 TYPE 81 ,1 ,IC(I) ,ID(I) ,EGY(I) ,B,YAA,ER(I) ,RATEUN

WRITE (4, 81 )I,IC(I) ,ID(I) ,EGY(I) ,B,YAA,ER(I) ,RATEUN

81 F0RMAT(1X,I4,I4,I5,F7.1 , F1 0 . 1

, F1 3 . 1, F7 . 2 , 1 X, 1 PE1 2 . 5

)

IF (SECS .LT. 1 . ) GO TO 40

TYPE 89 , RATESC, RATETO

WRITE(4 ,89 ) RATESC, RATETO

89 FORMAT (/IX, 'DECAY CORRECTED RATE = ',F10.3,' COUNTING

& RATE AT TO = ' ,1 PE12.5/)

D-14
40 CONTINUE

TYPE 9 ! LINE FEED

TYPE 3

WRITE(4 ,3)

WRITE (4,50)

50 FORMAT (/, 1 5X DETAILS REGARDING THE BACKGROUND SUBTRACTION'/)

WRITE (4,55)

WRITE (4,56)

55 FORMAT ( 2X, ' PEAK AREA LOW BKG HIGH BKG'

&,' BLOW BHIGH BL/BH BKG/T % ERROR')

56 F0RMAT(2X,' LIMITS LIMITS LIMITS')

WRITE (4,9)

57 F0RMAT(2X,I2,1 X,I4, '-' ,I4,1X,I4,'-',I4,1X,I4,'-',I4,

&1X,F7.1 ,1X,F7.1 ,1X,F7.3,1X,F7.4,1X,F7.2)

DO 58 1=1 ,NUM

D WRITE (4,59) CPK , IC ( I )

, ID( I ) ,NL( I ), NR ( I

)

D59 FORMAT (5X, F9 . 2 , 1 4 , 1 4 , 12 , 12

)

I1=IWL(I)-NL(I)+1

I2=IWH(I)+NR(I)-1

Z1=BL0W(I)/BHIGH(I)

Z2=BKG(I)/T0TAL(I)

58 WRITE(4 ,57)1 ,IC(I) ,ID(I) ,11 , IWL( I )

, IWH( I ), 12 ,BLOW( I )

,

&BHIGH(I),Z1 ,Z2,ER(I)

WRITE (4,9) ! LINE FEED

WRITE (4,3)

WRITE(4 ,3)

52 WRITE(4,9)

TYPE 9 ! LINE FEED

RETURN ! TO MAIN PROGRAM

END ! OF SUBROUTINE EPEAK

C

C

D-15

Appendix E

C PROGRAM SCATFO TO CALCULATE SCATTERING CORRECTION FOR ACTIVATION

C FOILS IN CAVITY FISSION SOURCE. DATA MAY BE ENTERED FROM A DISK FILE

C OR FROM THE KEYBOARD

CHARACTER*^ ANAME(50)

DIMENSION GRAD(50)

DIMENSION IT(50) ,P0S(50) ,TH(50)

OPEN (3 ,STATUS='NEW'

)

TYPE 97

97 FORMATC 1 X, ' ENTER FILE NUMBER FOR DATA; 5 FOR KEYBOARD')

READ(5 ,100)NF

100 FORMAT(IIO)

IF(NF.NE.5)G0 TO 101

TYPE 99

99 FORMAT (IX, 'ENTER NUMBER OF FOILS, C.R.')

GO TO 104

101 OPEN(NF,STATUS='OLD'

)

104 READ(NF,100) N IF(NF.NE.5)G0 TO 103

TYPE 98

98 FORMATd X, 'ENTER FOIL ID IN A4 FORMAT FOLLOWED BY TYPE,'

8=,/' POSITION IN MILS, THICKNESS IN MILS, FOLLOWED BY'/

&' C.R. AND THEN NEXT FOIL. USE 1 FOR IN, 2 FOR NI AND 3 FOR AL '

)

103 READ(NF,1 10) (ANAME(J) ,IT(J) ,POS(J) ,TH(J) ,J=1 ,N)

WRITE (7 ,1 1 1 ) (ANAME(J) ,IT(J) ,POS(J) ,TH(J) ,J=1 ,N)

1 10 FORMAT(A4,I10,2F10. 1

)

111 FORMAT(1X,A4,nO,2F10.1 ) WRITE(3,96)

96 FORMAT (2X, ' ID' ,6X, 'GRAD. ' ,8X, ' TYPE ' ,2X, ' POSITION' ,2X, 'THICKNESS

& FOIL SCAT MUSC '

)

DO 1000 1=1 ,N

GRAD(I)=1+21 .2E-6*POS(I)*POS(I)

SCAT=0

IF(IT(I )-2)400 ,500 ,600

400 DO 450 J=1 ,N

IF(I.EQ.J)GO TO 450

X=ABS(POS(J)-POS(I)

)

Y=-.001 +( .0054*EXP(-.0728*X)

)

Y=Y*TH(J)/10.

120 FORMATd X,E10. 3)

SCAT=SCAT+Y

450 CONTINUE

ALSCAT=SCAT+.0046+( .00255*TH(I )/5. ) GO TO 800

500 DO 550 J=1 ,N

IF(I.EQ.J)GO TO 550

X=ABS(POS(J)-POS(I))

C

CC

C

C

C

INPUT DATA:

ANAME(J)= FOIL ID

N = THE NUMBER OF FOILS

IT(J) = FOIL TYPE

POS(J)= POSITION OF FOIL IN MILS

TH(J) = THICKNESS OF FOIL IN MILS

E-1
Y=-.001 7+(.00356*EXP(-.08l5*X)

)

Y=Y*TH(J)/10.

SCAT=SCAT+Y

550 CONTINUE

ALSCAT=SCAT-.002+( .001 7*TH(I)/10. ) GO TO 800

600 DO 650 J = 1 ,N

IF(I.EQ.J)GO TO 650

X=ABS(POS(J)-POS(I))

Y=-.oo2+( .oo2*EXP(-. ^^^*x))

Y=Y*TH(J)/10.

SCAT=SCAT+Y

650 CONTINUE

ALSCAT=SCAT-.01 1 GO TO 800

800 WRITE(7,102)ANAME(I) ,GRAD(I),IT(I) ,POS(I) ,TH(I),SCAT,ALSCAT

WRITE(3,102)ANAME(I) ,GRAD(I) ,IT(I) ,POS(I) ,TH(I) ,SCAT,ALSCAT

102 FORMAT(1X,A4,F10.4,I10,2F10.1 ,3X,2E12.3)

1000 CONTINUE

CL0SE(3)

IF (NF.NE.5)CL0SE(NF)

STOP

END

E-2
Appendix F

10

20

30

40

50

60

70

80

90

1 00

1 10

120

125

126

1 30

140

150

160

170

180

190

200

REM SAT260

THE DATA

SEQUENCE:

X 1 ( I

...11 /21 /83

VARIABLES IN

COUNTER

C (I ) . . . X3 (I ) . . T (I ) . . . X5 ( I ) . . X6 ( I ) . .

' X 7 ( I) . REM Fl* IS

REM NAME

PRINTER IS 701,30

PRINT CHR* (27) ?<'?<! IL"

CLEAR

DIM C ( 260 ) , X 1 ( 260 ) , X5 ( 260

)

DISP "DATA FILE TO BE USED

INPUT Fl$

FIRST

Cu

Cqc

5ec*32

DAY COUNTER

hhmmsB

-TIME INTERVAL

THE DATA FILE

X3 ( 260 ) , X7 ( 260 ) , X6 ( 260 ) , T ( 260

)

ASSIGNtt

READ# 1 FOR 1=1

READ# 1 210 NEXT I 220 ASSIGNtt

1 TO Fl$

1 ; N

TO N

I + 1 ; X 1 ( I ) , C (I ) X3 ( I ) , T ( I ) , X5 ( I) , X6 ( I ) , X7 ( I)

230

240

250

260

270

280

300

310

320

330

340

350

360

380

410

420

430

440

1 TO * T ( 1 ) , X 1 ( 1 ) , C ( 1 ) , X3 ( 1 ) , X7 ( 1) =0

PRINT "SAT260 DATA ANALYSIS"

PRINT " 11/83 version"

PRINT

PR I NT

PRINT

PRINT

PRINT

FOR I PRINT

IMAGE 3D,2X,7D,3X

NEXT I

"N = ";N

"DATA VERIFICATION FOR

CI Cu

:1 TO N

USING 340

Fl*

Cqc hhmmsi

I , X 1 ( I ) , C ( I ) , 7D,3X,7D,5X

X3 ( I ) , X6 (I ) , T ( I)

6D, 3X , 60, 3X , 6D

sec*32

7(1)

HPsec"

PRINT

FOR 1=1 TO N

NEXT I PRINT

PR I NT

Al , A0=X5 ( 1

)

450 FOR 1=2 TO N

460 IF X5(I)<>A1

470 NEXT I 480 PRINT USING

490 IMAGE "DATA

500 PRINT USING

510 IMAGE " 520 PRINT

530 DISP

540 PRINT

550 DISP

THEN A1=X5(I)

490 : GOES

510 :

AO

FROM

Al

TO

DAY" , 4D

DAY" , 4D

"DIFININ6 THE IRRADIATION PERIOD'

"DEFINING THE IRRADIATION PERIOD"

560 DISP

570 PRINT

580 DISP "WITH

590 DISP "LIKE

600 INPUT Z5

610 IF Z5>1 THEN 630

F-1

WHICH INTERVAL WOULD YOU"

ANALYSIS TO START ";
(b3U

640

650

660

670

680

NOT

690

700

710

720

730

740

750

760

770

780

790

800

810

820

830

840

350

860

870

880

890

900

9 1

915

916

920

2 1 8<:: 2 1 90

220(

22 IC

2240

225(:

226(:

227<:

228(:

2290

2300

23 IC

233C

234(1

'-\ -T cr /- ^ J* ^ vJ 236(:

2370

238(;

239(:

240<:

24 1<: 242<:

Q5=Z5 - 1 DI3P "WITH WHICH INTERVAL DO YOU WANT

DISP "ANALYSIS TO CONCLUDE";

INPUT N

T(Q5) , XI (Q5) ,C(Q5) , X3(Q5) , X7<Q5)=0

Z0=N--Z5+1 li' REM ZO IS CORRECT DIVISOR

1 "ANALYSIS INCLUDES INTERVAL: ";Z

" THROUGH I NTERVAL : " ; N

FOR AVERAGES INSTEAD OF N, WHEN ZO 1

O.I. :";X6(Z5-

O.I. :";X6(N)

1

PR I NT

PR I NT

PRINT

PRINT "E.S.T. AT S.

PRINT "E.S.T. AT E,

PRINT

X1=N

T=0

T1=0

CI , S1=0

FOR I =05 TO N

T(I)=T<I) /32

Tl=Tl-i-T ( I ) S 1 =S l+C ( I

Z1 = X6(I) *10 ••(-4)

X6 ( I

) =IP (Zl ) *3600

Z1=FP (Zl ) *100

X6 (I ) =X6 ( I ) +IP ( Z 1 ) *60+FP ( Z 1 ) * 100

X6 ( I

) =X6 ( I ) +X5 ( I ) *86400

IF I=Q5 THEN 900

Cl=Cl-(-C ( I ) /T ( I ) * (X6 ( I) -X6 (I-l ) ) NEXT I T=X6 (N) -X6 (Q5)

PRINT "IRRADIATION TIME :";T;"s"

PRINT

C1=C1/T

PRINT

"THE CORRECTION IS THE RATIO OF"

"ACTUAL ACTIVITY AT EOI TO THAT"

"WHICH WOULD HAVE BEEN PRODUCED"

"BY THE SAME FLUENCE AT CONSTANT"

"POWER" ft' PRINT

PRINT

PRINT

PRINT

PRINT

PRINT

CLEAR

DISP

DISP

DISP

DISP

DISP

DISP

DISP

DISP

DISP

DISP

DISP

DISP

INPUT

DISP

"WHICH LAMBDA DO YOU WANT'

"4.

"6.

"8.

"9.

" 10

A

"WAIT"

NUCLIDE

In

Sc-46

Ni

Mn-56

Mn-54

Au

Na

Sc-47

Sc-48

Ba-La

1 1/2

4. 486h

83. 85d

70. 81d

2. 576h

3 1 2 . 5d

2. 696d

15. OOh

3. 40d

43. 8h

12. 789d

4. 292E-5"

9. 568E-8"

1 . 1330E-7"

7. 4744E-5"

2. 567E-8"

2. 9757E-6"

1 . 2836E-5"

2. 3596E-6"

4. 3959E-6"

6. 2373E-7"

ON A GOTO 2390,2400,2410,2420,2430,2440,2450,2460,2470,2475

L=. 00004292 li' L$="In" i? GOTO 2480

L=. 00000009568 i*' L*="Sc-46" i? GOTO 2480

L=. 0000001 133 I*' L*="Ni" li' GOTO 2480

L=. 000074744 i? L* = "Mn-46" !»• GOTO 2480

F-2
243

24 4(

24^

246(

247(

247^

248'.

249(

254(

2560

257(

25S(

259'.

260(

261'

262(

26.3(

264(

26^

266(

267<:

2680

269!:

270

C

271(

273

274(

2750

276C

277C

L=. 00000002567 i? L*="Nn-54" H' GOTO 24S0

L=. 0000029757 i? L*="Au" li' GOTO 2480

L=. 000012836 \l' L$="Na" li' GOTO 2480

L=. 0000023596 li' L*="Sc-47" i? GOTO 2480

L=. 0000043959 5' L*="Sc-48" i? GOTO 2480

L=. 0000006273 li' L*="Ba-La"

33, Dl , S2=0

FOR J=Q5+1 TO N

C=L* ( X6 ( J ) -- X6 ( J-1 ) ) S2=S2+C (J ) /T ( J ) * (1-EXP (-C) ) /L*EXP <- (L* ' T- ( X6 ( J

) -X6 (Q^

S3=S3+1* (1-EXP (-C) ) /L*EXP (- (L* (T- ( X6 (J) -X6 (OS) ) ) )

)

NEXT J

S=L*B2/C1

D=L*33

C3= 1 -EXP (- (L*.T) ) C2=S/C3

PRINT

PRINT " PRINT li' PRINT

PRINT "NUCLIDE ... " ; L*

PRINT "LAMBDA = ";L;" 1/SEC"

PRINT li' PRINT li' PRINT

PRINT USING 2670 ; S

IMAGE "SATFAC = ",D.DDDDe

PRINT USING 2690 ; 0

IMAGE "FLAT CASE SFAC = "D.DDDDe

PRINT USING 2710 ; C3

IMAGE "XACT FLAT CASE SFAC = ", D.DDDDe

PRINT USING 2730 ; C2

IMAGE "CORRECTION = ", D.DDDDe

PRINT li' PRINT li' PRINT li' PRINT

CLEAR

GOTO 2240

END

) ) ) ) )

F-3

Appendix G

TEST REPORT FOR NEUTRON FLUENCE STANDARDS

PRODUCED BY AN IRRADIATION AT THE NBS

2 3 S U CAVITY FISSION SOURCE

Irradiation No. : 8601 Report Date: March 6, 1986

Sensor Materials Irradiated: Nickel, Gold, and Scandium

(designations: Au-1,2; Sc-S-10; Ni-3,*l,5,6)

End of Irradiation (EOI): Jan 20, 1986 at 2137:00 EST (Day 20 of 1986)

REPORT FOR; Ni-5

Free Field Fission Neutron Fluence: 4> = 8.266 x 10'^ n/cra^ ± 2 % (1 o)

Physical Description of Material: Mass = 0.2856^1 g

Additional Parameters of Irradiation:

Decay Correction Factor: C = 0.9881

Length of Irradiation: T = 210960 s Departure from free-field activity due to Neutron Scattering (p ):

Other Radioactivities:

Since the sensors are natural metal samples or in some cases special

alloys, activities from neutron reactions in other isotopes and

impurities may be present. Ordinarily, this will not be a problem

for Ge(Li) and other high resolution gamma detectors.

The following lines have been observed above the annihilation

energy:

Form: 12.7 mm dia. X 0.25 mm thk disks

Type: Natural element nickel

sc

= -(0.7 ± 0.3) %

6 0 Co < 0.1 % of '^Co activity on 1/28/86

This fluence standard prepared by: For the Director

G.P. Lamaze

Ionizing Radiation Division

G-1

R.S. Caswell, Chief

Ionizing Radiation Division
Relationship of Teat Report Data to the Measured Activity at End of

Irradiation (EOI ) for Cavity Fission Neutron Source Irradiations

The neutron fluence or fluence rate standard has been exposed to a fission

neutron fluence that is perturbed by a small cavity return flux and by the

effects of neutron scattering in fission source disks, encapsulations, and

the fluence standards themselves. This test report certifies a free-field

fission neutron fluence or fluence rate, i.e. corrected for neutron

scattering. This correction, based on calculation, is reported as the

fractional departure from the free-field activation attributable to neutron

scattering including cavity return. On this basis the total activity

measured at EOI is

where: A = total activity measured at EOI m

A = net free-field activity at EOI

y^^ = fractional departure from free-field activity due to

neutron scattering. (A minus sign means that the observed

activity is lower than it would be in free-field conditions

because of a net outscatter of neutrons.

)

Activation equation for neutron fluence . For simple decay, the unperturbed

activity of the neutron fluence standard in disintegrations per second is

related to the free-field fission neutron fluence by the activation equation — = a • * (2)

ACN

where: A = net free-field activity at EOI (dps)

X = decay constant of reaction product

C = decay correction factor given in test report. For an uninterrupted irradiation of length T at a constant fluence

rate, C is equal to [ (

1 -exp( -XT) ) /XT] . (When it is appropriate

to report average neutron fluence rate, the decay correction

factor is given as the product XTC — see below.)

N = number of reaction isotope atoms in the fluence standard

0 = fission-spectrum-averaged reaction cross section (cm2)

G-2
2

* = free-field fission neutron fluence (n/cm ) given in this test

report. (Note: the symbol * is recommended by ICRU and ASTM

E170.

)

= reaction probability

Activation equation for average neutron fluence rate. For decay

half-lives short compared to the length of irradiation, the activation

equation is commonly expressed in terms of the average reaction rate

(experimentally, the saturated specific activity): — = 0 • <<|)> (3)

Catc]n

[ATC]N

" saturated specific activity

[ATC] = decay correction factor for average neutron fluence rate. For an uninterrupted irradiation of length T at constant fluence rate,

[ATC] is equal to [l-e"'^'^].

T = length of irradiation given in the test report.

<<J)> = average, free-field fission neutron fluence rate (n/cm^s) given

in this test report.

o<(|)> = average reaction rate

Activation equation for fission product activity. When the observed

gamma disintegration is from direct formation of a fission product,

Equation 2 becomes

A = Y'A 'C-N* (a-<I>) , (^4

)

where Y is the chain fission yield. When the fission product actvity is

from the daughter of a radioactive parent with a half life longer than

that of the daughter, the appropriate free-field activity at equilibrium

is

A = [A'/(A'-A)]-Y-A-C-N-(a-<I>) (5)

where A is the decay constant of the parent activity and X' that of the

daughter.

G-3
Recommended Procedure for Establishing a Cal ibration Factor ( X2 s

)

Use Equation (1) and either Equation (2) or Equation (3) to derive a free-field neutron fluence or average fluence rate, respectively, from the

measurement of a particular activity of the neutron sensor. Define a calibration factor for this activity as the ratio of the NBS certified

fission neutron fluence or fluence rate to the derived neutron fluence or fluence rate. Then, to obtain an NBS calibrated result for subsequent

measurements with this activity, either multiply observed reaction

probability by the calibration factor or divide the detector cross section

by the calibration factor. This calibration procedure will be most

effective if the derived fluence or fluence rate for the NBS Standard is

obtained with the same reaction parameters, and with the same fission

spectrum shape and detector cross section set as are used in the transport

calculations which provide spectrum-averaged detector cross-sections for

subsequent neutron field measurements.

Revised: November 3, 1986

G-4
ASSIGNMENT OF UNCERTAINTIES ( ± 1 a)

Plus Transfer Measurements from the ^^^cf standard Fission

Neutron Field to the ^ ^Sy Cavity Fission Source :

Percent

(1) Absolute neutron source strength of the 252cf source 0.9 S+R

(2) Distance measurement 0.6 R (3) 252cf to 235u spectrum-averaged cross section ratio

for Pu(n,f) monitor 0.1 S (4) Correction for neutron scattering for Pu(n,f) 0.9 S (5) Reproducibility of flux transfer procedure (Including

counting statistics, and electronic system stability

over period of measurement) 0.4 R

Subtotal (summed in quadrature) = 1.5

o Preparation of Neutron Fluence Standards at the

235|j Cavity Fission Source :

(1) Gradient corrections and positioning 0.4 S+K

(2) Corrections for neutron scattering for ii5in(n,n') 1.2 R (3) Reproducibility of fluence monitoring including random

summing corrections and uncertainties in activation

decay corrections 0.4 R

Subtotal (summed In quadrature) =1.3

TOTAL UNCERTAINTY = 2.0

*

S is for systematic uncertainty and R is for random uncertainty.

G-5
NBS-n4A (REV. 2-ec) U.S. DEPT. OF COMM.

BIBLIOGRAPHIC DATA

SHEET (See instructions)

1. PUBLICATION OR

REPORT NO.

NBS/SP-250/14

2. Performing Organ. Report No, 3. Publication Date March 1988

4. TITLE AND SUBTITLE

NBS MEASUREMENT SERVICES:

ACTIVATION FOIL IRRADIATION BY REACTOR CAVITY FISSION SOURCES

5. AUTHOR(S)

G. P. Lamaze and J. A Grundl

6. PERFORMING ORGANIZATION (If joint or other than NBS. see instructions)

NATIONAL BUREAU OF STANDARDS

U.S. DEPARTMENT OF COMMERCE

GAITHERSBURG, MD 20899

7. Contract/Grant No.

8. Type of Report & Period Covered

Final

9. SPONSORING ORGANIZATION NAME AND COMPLETE ADDRESS (Street. City. State. ZIP)

Same as item 6

.

10. SUPPLEMENTARY NOTES

Library of Congress Catalog Card Number 88-600513

I I Document describes a computer program; SF-185, FlPS Software Summary, is attached.

11. ABSTRACT (A 200-word or /ess factual summary of most significant information,

bibliography or literature survey, mention it here)

If document includes a significant

This document describes the National Bureau of Standards (NBS)

calibration service 44090C (formerly 8.1R) which operates in the

following way: (i) dosimetry sensors (metal foils, nuclear track

detectors, wires, crystals, etc., supplied by the customer or by

NBS) are irradiated to a measured neutron fluence in a fissionspectrum neutron field (for short lived reaction products, the

samples are irradiated to near saturation and the fluence rate

specified); and (ii) the irradiated dosimetry sensors are shipped

to the customer, followed by a test report stating the fluences or fluence rates and associated uncertainties. The scope and

philosophy of the service, the neutron field characteristics,

irradiation procedures, and the uncertainties in the reported

neutron fluences or fluence rates are discussed. Typical maximum

fluences are of the order of 3 x 10^^ neutrons/cm^ , with combined

(la) uncertainties of ± 2%.

12. KEY WORDS (Six to twelve entries; alphabetical order; capitalize only proper names; and separate key words by semicolon s) activation; dosimetry; fast neutrons; fission spectrum; neutron fluence

1 13. AVAILABILITY

Unl imited

I I For Official Distribution. Do Not Release to NTIS

Order From Superintendent of Documents, U.S. Government Printing Office, Washington, D.C. 20402.

Order From National Technical Information Service (NTIS), Springfield, VA. 22161

14. NO. OF

PRINTED PAGES

56

15. Price

*U.S. Government Printing Office : 1988 - 201-597/825'i8 USCOMM-DC 6043-P80


THE SP 250 SERIES ON NBS MEASUREMENT SERVICES*

SP 250-1 Spectral Radiance Calibrations 003-003-02792 $3.50

SP 250-2 Far Ultraviolet Detector Standards 003-003-02810-0 $4.25

SP 250-3 Radiometric Standards In ttie Vacuum

Ultraviolet 003-003-02806-1 $6.50

SP 250-4 Fricke Dosimetry in High-Energy Electron Beams

003-003-02816-9 $2.75

SP 250-5 Alpha-Partlde Calibrations SN003-003-0283-1 $2.00

SP 250-6 Regular Spectral Transmlttance 003-003-02805-3 $3.25

SP 250-7 Radiance Temperature Calibrations 003-003-02827-4 $2.25

SP 250-6 Spectral Reflectance 003-003-02812-6 $7.00

SP 250-9 Calibration of Beta-Particle-Emitting

Ophthalmic Applicators

003-003-0281 7-7 $2.00

SP 250-10 Radioactlvtty Calibrations with the "Air" Gamma Ionization Chamber, and Other

Radioactivity Calibration Capabilities

003-00^-02824-0 $2.25

SP 250-1 1 Dosimetry for High-Dose Applications

SP 250-12 Neutron Personnel Dosimetry

003-003-02811-8 $2.50

SP 250-13 Activation Foil Irradiation with Californium

Fission Sources

SP 250-14 Activation Foil Irradiation by Reactor Cavity

Fission Sources

SP 250-15 Photometric Calibrations SN003-003-02835-5 $4.25

SP 250-16 Calibration of X-Ray and Gamma-Ray

Measuring instoments

SP 250-17 The NBS Photodetector Spectral Response

Calibration Transfer Program

SP 250-18 Neutron Source Strength Calibrations

SP 250-19 Calibration of Gamma-Ray-Emitting

Brachytherapy Sources

SP 250-20 Spectral irradlance Calibrations 003-003-02829-1 $5.50

SP 250-21 Calibration of Beta-Partlcle Radiation Instrumentation

SP 250-22 Platinum Resistance Thermometer

Calibrations 003-003-02831-2 $17.00

SP 250-23 Uquld-in-GiassThemTOmeter

Calibration Sen/lce

SP 250-24 Standard Ceil Calibrations 003-00:H)2825-8 $2.75

SP 250-25 Calibration Sen^ for inductive Voltage Dividers

SP 250-26 Phase Angle Calibrations

SP 250-27 AC-DC Difference Calibrations

SP 250-28 Solid-state DC Vott^ Standard Calibrations SN003-003-02842-8 $2.00

SP 250-29 Traceable Frequency Calibrations SN003-003-02844-4 $2.25

SP 250-30 GOES Satellite Time Code Dissemination:

Description and Operation

SN003-003-02845-2 $2.75

SP 250-31 Mass Calibrations at the National Bureau

of Standards

•Those entries containing a stod< number (003-003 ) and price can be purchased from the Superintendent of Documents, U.S. Government Printing Offtee, Washington, DC 20402. GPO will accept chedo, money orders, VISA, and Mastercharge. For more Information, or to place an order, call (202)783-3238. Be sure to use the stodc number in all orders. Titles without stock numbers are in preparation.
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Penalty fcx Private Use $300

Sttmulating America s Progress

1913-1988