NBS Measurement Services:

Activation Foil

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, conpiles, 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 quantum 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 instrunentation for

radiation programs and facilities; provides national leadership

in the standardization of nuclear instrunentation; and develops

new competence In radiation sources and Instrumentation. The

Division consists of the following Groups:

• Accelerator Research • Linac Operations • Electronic Instrumentation

• 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 WITH

CALIFORNIUM FISSION SOURCES

George P. Lamaze

James A. Grundl

Center for Radiation Research

National Measurement Laboratory j National Bureau of Standards

Gaithersburg, MD 20899

I

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: 88-600508

National Bureau of Standards Special Publication 250-13

Natl. Bur. Stand. (U.S.), Spec. Publ. 250-13, 38 pages (Mar. 1988)

CODEN: XNBSAV

Commercial products— materials and instruments -are Identified in tfiis document for the sole pur- pose of adequately describing experimental or test procedures. In no event does such identification imply recommendation or endorsement by the National Bureau of Standards of a particular product; nor does It imply that a named material or instrument is necessarily the best available for the purpose

it serves.

U.S. GOVERNMENT PRINTING OFFICE

WASHINGTON: 1988

For sale by the Superintendent of Documents, U.S. Govemment 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 250-13), NBS Measurements Services: Activation Foil

Irradiation with Californium Fission Sources, by G. P. Lamaze and J. A.

Grundl, is the thirteenth 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 44080C 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

Acting Chief

Measurement Services

Chris E. Kuyatt

Di rector

Center for Radiation Research
ABSTRACT

This document describes the NBS calibration service 440800

(formerly 8.1Q), 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 certified neutron fluence in a 252Qf spontaneous fission 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 1 x iQis

neutrons/cm2 and maximum fluence rates of 1 x 10^ neutrons/cm^ sec,

with combined (la) uncertainties of ± 1.2%.

Key words: activation; californium; dosimetry; fast neutrons; fission

spectrum; neutron fluence; neutron fluence rate

i V
TABLE OF CONTENTS

I. DESCRIPTION OF SERVICES AND FACILITIES 1

I. a. Generic Description 2 I.b. Available Fission Neutron Exposures 2 I.e. Corrections for Neutron Scattering at 5 cm Position 2

I. d. Configuration and Characteristics 2

II. NEUTRON FIELD CHARACTERIZATION 8

II. a. Neutron Fluence 9

II. b. Neutron Spectrum 10

Il.b.l. Measurement 10

II. b. 2. Evaluated Spectrum 10

III. DETAILED OPERATING PROCEDURES 12

IV. FLUENCE CALCULATION AND UNCERTAINTIES 13

IV. a. Fluence and Fluence Rate Calculations 13

IV. b. Uncertainties 14

V. BENCHMARK REFERENCING 15

VI. SAFETY 16

REFERENCES 17

APPENDIX A. Listing of CFYLD A-1

APPENDIX B. Listing of CFGFAC B-1

APPENDIX C. Sample Test Reports C-1
LIST OF TABLES

Page

TABLE 1. 252cf Fission Neutron Field Parameters and Uncertainty

Components 5

TABLE 2, Typical Neutron Fluence and Sensor Response Perturbations

due to Neutron Scattering 8

TABLE 3. Evaluated Fission Neutron Spectra for 252cf 235u -jn

45-Group Format 11

TABLE 4. Uncertainty Estimates for 252cf ^pcl 235u Evaluated Fission

Neutron Spectra 12

TABLE 5. Net Scattering Corrections for some Common Sensors 14

vi
LIST OF FIGURES

Page

Fig. 1 The NBS evaluation of the 252Qf fission neutron spectrum 3

Fig. 2 Typical arrangement of six activation sensors around the

252cf source 4

Fig. 3 Scattering materials in the low-mass 252cf fission neutron source capsule 6

vii

I. DESCRIPTION OF SERVICES AND FACILITIES

NBS maintains standard fission neutron fields to supply calibration

irradiations and certified fluence and fluence rate standards. The NBS 252cf

Fission Neutron Source (typical maximum fluence of 1.0 x 10^^ neutrons/cm^ in

24 hours) provides one such standard neutron field spectrum. Known fluences

and fluence rates are based upon Cf source strengths that are traceable to

NBS-I, the U.S. national fast-neutron source standard.

This calibration service (44080C) operates in the following way:

1) dosimetry sensors (metal foils, nuclear track detectors, wires, crystals,

etc., supplied by the customer or by NBS) are irradiated to a certified

neutron fluence in a 252Qf spontaneous fission neutron field (for short lived

reaction products, the samples are irradiated to near saturation and the

fluence rate specified), and 2) the irradiated dosimetry sensors are shipped

to the customer, followed by a test report stating the fluences and associated

uncertainties. For a typical activation sensor, the customer measures the

activity of the irradiated sensor with his own sensor 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 free-field 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 = A • C • N • a • $ (1)

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

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

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

(this corrects for activity lost during the irradiation),

N = number of atoms of 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

ICTN = • *

where: ttTKT = reaction rate probability (disintegrations per second per

nucl eus )

,

(j) = ($/T), the effective neutron fluence rate (n/cm^ s) or flux

density, and

T = the length of the irradiation given in the test report.

1
I. a. GENERIC DESCRIPTION

A standard 252(;f neutron field consists of neutrons from the spontaneous

fission of 252Qf y^j^h little or no energy degradation. The median energy of

the spectrum is 1.68 MeV, with 98% of the neutrons between 0.1 MeV and 8 MeV

(see fig. 1). Neutron fluence rates in the range of 10'^ n/(cm2 s) are obtained in isolated environments near small, intense ^^^Cf fission sources. Neutron fluences are established in terms of neutron source strength,

irradiation time, and source-sensor distance; no microscopic nuclear data or irradiation monitors are required. Certified free-field fluences of up to

10^3 n/cm2 may be obtained with uncertainties as low as ± 1.2% (la).

Measurements of the 252Qf fission neutron spectrum and the closely

related the ^ssy fission neutron spectrum are extensive and wel 1 -documented

[1]. These two standard neutron fields are therefore much better known than

any other benchmark employed for reactor dosimetry calibration. Moreover, in

the energy range above 2 MeV, many neutron fields in and around test and power

reactors have fission-spectrum-like components.

I.b. AVAILABLE FISSION NEUTRON EXPOSURES

e fluence rate (5 cm from source) ~ 2 x 10'^ n/(cm2 s)

© nominal maximum fluence 1 x lO^^ n/cm2

9 accuracy of free-field fluence (la) ± 1.2%

I.e. CORRECTIONS FOR NEUTRON SCATTERING AT 5 CM POSITION

o Net perturbation for threshold sensor reaction rates: < (1.0 ± 0.8)%

(la)

• Net perturbation for ^ssy (n,f) sensor reaction rate: (1.2 ± 0.4)%

(la)

I.d. CONFIGURATION AND CHARACTERISTICS [2,3]

The two irradiation locations at NBS are distinguished by the degree to

which the neutron source is isolated from environmental neutron return:

Location A: Large room with thick walls and open ceiling. Source 2.2

meters above floor, nearest wall 4.1 meters.

Location B: Corner area of room with thin walls and ceiling. Source 2.8

meters above floor, nearest wall 4.3 meters.

At both locations a lightweight source-sensor assembly is available for

irradiation of passive and active neutron sensors. The source-sensor assembly

with six activation sensors mounted on three axes is shown in figure 2.

Location A is currently being used for calibration of health physics type

dosimeters. All activation foil irradiations are now done at location B. All

further site specific properties discussed in this report will apply to

location B only, although the site specific properties are very similar.

2
CS/2->«-*tN3/N> C3> IHd

3

The californium source capsule, shown in fiyure 3, is made up of a diskshaped 252Qf deposit in an aluminum pellet encapsulated in a thin-walled

stainless steel cylinder. The position of the californium deposit relative to

the capsule surface is known to a maximum displacement ± 0.5 mm based on constraints of fabrication and x-ray photographs. Current capsule designs

feature a short attachment stem in place of the first few cm of the source

guide tube; the 252cf deposit enclosure is unchanged.

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 platinuniridium capsules of 0.2 mm wall thickness. This source has been calibrated by

several independent methods [5-7] and the source strength 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 [8]. The uncertainty of the source strength obtained in this

fashion is ± 1.1% (la)

.

TABLE 1. 252cf FISSION NEUTRON FIELD PARAMETERS AND UNCERTAINTY COMPONENTS

Free-field fission neutron fluence rate

(5-cm distance; source 6 x 10^ n/s) 2 x 10^ n/(cm2 s]

Source decay rate 2.2% / month

Typical maximum fluence lO^^ n/cm^

Source capsule and support scattering correction < 1.3%

(inelastic plus net elastic inscatter) (see Table 2)

it Gamma-ray exposure at 5 cm ~ 170 R/h

(2.8 yr after separation)

Uncertainties (la) in the free-field fission neutron fluence due to :

Source strength ± 1.1%

Source capsule and support scattering ± 0.4%

Distance measurements ± 0.2%

(typical for compensated-beam geometry)

Quadrature sum: ± 1.2%

*This estimate does not include x-rays which the light-weight encapsulation

may not absorb.

Neutron field parameters and uncertainty estimates for a nominal 5 cm source-to-sensor distance are given in Table 1. The source strength

uncertainty of ± 1.1% dominates the composite uncertainty of ± 1.2% for the

5
7.9mm

1.7 mm

-0.50mm

7.6 mm

SOURCE

GUIDE

TUBE

STAINLESS STEEL

CAPSULE LID (0.59g)

Cf DEPOSIT (-3mg)

OgSO^ DISI

1.4 mm^

)

(CfgOgSO^ DISPLACEMENT

ALUMINUM PELLET

( 6.4 mm dio. x 5.7 mm

LONG : 0.48g )

STAINLESS STEEL CAPSULE

(TYPE 304 i 76mm dio -, 0.80g)

Fig. 3 Scattering materials in the low-mass 252cf fission neutron source capsule.
free-field fission neutron fluence. The irradiation geometry shown in

figure 2 is termed compensated-beam geometry and refers to the experimental

practice of placing sensors of similar sensitivity in pairs on opposite sides

of the source, and nearly equidistant from it. The geometric mean of the

responses of the two sensors in this case can be expressed in terms of a mean neutron fluence which is a function of sensor separation with very little

dependence on source position.

It can be shown that the percent uncertainty due to source displacement

on axis with the sensors is (Ar2/r2-Ar2) x lOO.

For a 0.5-mm displacement where r = 50 mm, the uncertainty is 0.01'/,. For

displacement perpendicular to the displacement axis, the maximum percent

decrease in the fluence rate is (2»[/r^+Ax-r]/r) x 100. For a perpendicular

displacement of 0.5 mm where r = 50 mm, the fluence rate is decreased by 0.1%.

The uncertainty in source position therefore does not contribute to the overall uncertainty. The principal percent uncertainty due to positioning can be

shown to be equal to (2Ad/d) x 100 where d is the measured distance between

samples and Ad is the uncertainty in measurement. For the Boice C786

coordinate measurement machine, the measuremnt accuracy for linear distances

is ± .005 mm (la). For our sample holders, which are not perfectly rigid, the

reproducibility of measurements is ± .04 mm (la). For the previously

discussed case of a sample 50 mm from the source (100 mm sample separation),

the uncertainty from positioning uncertainty is ± .08% (la).

Scattering Corrections . Neutron fluence perturbations attributable to room return and scattering in the source capsule and support structures are given in Table 2. Corresponding sensor perturbations are listed for a threshold and a low-energy response neutron sensor. The room-return component

is estimated from calculations matched to response-versus-di stance measurements in which the source is moved while the sensors are kept at their normal

5 cm position [9]. For sensors with low-energy response, a large cadmium

basket is placed around the source-sensor assembly to absorb room-return

thermal neutrons. The fission neutron return from this basket is less than

0.1%.

Three types of calculations are employed for estimating free-field

neutron fluences: (1) discrete ordinates calculation of a spherical cavity in

concrete to obtain the albedo from boundaries; (2) geometry and simple energy

degradation factors calculated for single-scatter events in individual pieces

to obtain corrections for scattering in source capsule and support structures;

and (3) calculations based on published analytical formulations involving a simple scattering kernel to obtain estimates of air scattering. Air scatter

contributions are less than 0.1% for source-sensor distances of 5 cm and

< 0.5% for distances up to 15 cm.

The degree of difficulty in correcting a free-field neutron fluence for

scattering in a sensor depends upon the mass, arrangement, and material of the sensor. The NBS has codes available that provide correction factors for

isotropic scattering of neutrons in lightly constructed sensors with cylindrical symmetry. More extensive Monte-Carlo calculations have been carried

out for special sensors, specifically the NBS double fission chamber.

7
TABLE 2. TYPICAL NEUTRON FLUENCE AND SENSOR RESPONSE PERTURBATIONS DUE TO

NEUTRON SCATTERING (5-cm DISTANCE)

Neutron fluence perturbations:

(above 0.4 eV)

room return 0.1%

source capsule scattering 0.8%

support structure scattering 0.4%

air scatter < 0.1%

Net perturbation of sensor response due to neutron scattering:^

235u(n,f) sensor with cadmium cover:

room return (0.1 ± 0.01)%

source capsule (0.8 ± 0.4)%

support structures (0.3 ± 0.1)%

238u(n,f) threshold sensor:

room return ^ (0.0 ± 0.01)%

source capsule (0.0 ± 0.5)%

support structures (0.3 ± 0.3)%

^All uncertainties are la.

Source capsule perturbations depend markedly upon sensor threshold.

Multiple scattering in more massive sensors are difficult to estimate and

often require auxiliary experiments. Generally, isolated 252cf fission

neutron fields are most appropriate for accurate, uncluttered exposures with

lightweight sensors.

II. NEUTRON FIELD CHARACTERIZATION

The energy spectrum of 252Qf spontaneous fission neutrons is similar to

that of 235u other fissionable materials. As such it provides a spectrum

of neutrons characteristic of the driving source for most of nuclear energy.

Because fission neutron spectra are similar and have been evaluated with the

same methods, the 252Qf spectrum is specified here together with the ^ssy

spectrum. A concise description of fission neutron spectra may be given in

terms of a broad energy range,

lower bound median upper bound

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

252cf 0.09 MeV 1.68 MeV 7.8 MeV

235U 0.08 1.57 7.2
where p = fraction of the neutron spectra above E ; and in a coarse seven- group display, (|)(E)aE, as follows: ^

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

(|)(E)dE ( 252cf) 0.047 0.184 0.220 0.194 0.200 0.146 0.009

(|)(E)dE (235U) 0.054 0.197 0.229 0.195 0.192 0.127 0.006

A more detailed description of the two spectra is given in Section II. b.

Prompt fission gamma-rays and gammas from fission products contribute

about equally to the total gamma-ray emission throughout the useful lifetime

of the source. As indicated in Table 1, the neutron-to-gamma ratio is about

1.1 X 105(n/cm2 s)per R/h.

II. a. NEUTRON FLUENCE (see also Section I.d)

The neutron emission rate for the singly-encapsulated sources employed at

NBS are ~ 6 x 10^ n/s when fabricated. The sources are periodically

calibrated in a flnSOi^ bath and the source strength is calculated for the day

of the irradiation (the decay rate is ~ 2.2% per month). For a neutron

fluence rate of ~ 2 x lO'' n/{cm^ s) at the 5-cm distance, the fluence for a one week exposure is ~ 1 x lO^^ n/cm^. Spatial gradients for uncoil ided

fission neutrons are just those of a point source in free space:

(A(j)/<j)) = (2Ar/r) Free-field fluence accuracy and sensor response perturbations

for typical irradiation arrangements are discussed in Section I.d.

In the case of compensated-beam geometry where nearly identical sensors

are placed on opposite sides of the source and nearly equidistant from it, a mean neutron fluence rate may be defined.

<$> = 4TT(r/2)2

S = source strength (n/s)

r = distance between sensors (cm),

which is proportional to the geometric mean of the sensor responses per atom;

« a a <$>

where D/N = sensor response per atom

a = reaction cross section

a = efficiency factor (same for both sensors).

[The exact form of the above expression departs from this approximation by

< 0.2% for D^/Nj within 20% of D2/N2.]

9
A small anisotropy results from the cylindrical shape of the fission source capsule. The ratio of the fluence rate averaged over all directions to

the fluence rate along a direction perpendicular to the capsule axis is 0.992

when the sensors do not subtend an angle of more than about 20° at the source.

II. b. NEUTRON SPECTRUM

Il.b.l. Measurement [10], [11]. Fission neutron spectra are most

accurately determined by laboratory measurement. (See, however, Refs. [12],

[13], and earlier work cited therein.) From 1952 to the present, welldocumented measurements of 252cf 235u fission neutron spectra have

employed many different techniques of neutron spectrometry [14]. Fission

spectra are thus the most widely studied fast-neutron energy distribution in

existence. The large body of documented fission spectrum data was evaluated

at NBS in 1975 [11]. The evaluation included an estimate of the spectrum

uncertainties in multigroup format based on the spread of the various data

sets. The 252Qf ^^d ^asy fission spectrum shapes recommended in the evaluation are based on the sixteen selected measurements listed in Table X-3 of

Ref. [1].

II. b. 2. Evaluated Spectrum . The NBS fission spectrum evaluation has

been chosen for presentation in this document. The 252cf spectrum is defined

up to 20 MeV by means of a reference Maxwellian, M(E), modified by four

piecewi se-conti nuous linear segments below 6 MeV plus one exponential segment

above 6 MeV, The reference Maxwellian for the 252cf fission spectrum is

M(E) = 0.667 /F . exp(-1.5E/2.13) , E in MeV,

with adjustment functions y(E) as follows:

Energy Interval (MeV) ypf(E)

0.0 - 0.25 1 + 1.20E - 0.237

0.25 - 0.8 1 - 0.14E + 0.098

0.8 - 1.5 1 + 0.024E - 0.0332

1.5 - 6.0 1 - 0.0006E + 0.0037

6.0 -20 1.0 exp[-0.03(E-6.0)]

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

tabulation, normalized to 1, of the evaluated 252cf fission spectrum is given

in Table 3 (and plotted in Figure 1) along with the corresponding ^ssy evaluated spectrum for comparison purposes. For other energy group structures this

tabulation may be interpolated as recommended in the footnote of Table 3.

10
TABLE 3. EVALUATED FISSION NEUTRON SPECTRA FOR 252cf AND 235u

IN 45-6R0UP FORMAT

252cf 2 35U

Lower

Energy

Boundary

(MeV)

Group

Fl uence

({) • aE

Cumul at! ve^^)

Fluence to Lower

Boundary

Group

Fl uence

4) • aE

Cumul at ! ve^^^

Fluence to Lower

Bounda ry

0,0 0.0039 1 0000 0.0048 1 .0000

0.05 0.0074 0.9961 0.0088 0.9952

0.1 0.0?1 9 0.9887 0.0249 0.9864

0.2 0.0140 0.9668 0.01 54 0.961 5 0 ?5 0.01 52 0 9528 0.01 66 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 0.871

0.0367 0.8582

0 6 0 0343 0.8374 0.0365 0.821 5 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 0.521

0.0536 0.491

5

1 .8 0.0484 0.4684 0.0487 0.4379

2.0 0.0442 0.4200 0.0441 0.3892

2.2 0.0206 0.3758 0.0204 0.3451

2.3 0.01 96 0.3552 0.01 93 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 0.1865 0.0290 0.1618

3.7 0.041

3

0.1 555 0.0378 0.1 328

4 2 0 0253 0.1 1 42 0.0227 0.09503

4 6 0 0200 0.0889 0.01 76 0.07233

5 n n 01 90 0 0689 0.01 64 0.05473

5 5 0.01 40 0.0499 0.01 18 0.03833

0*01 0? 0 0359 0.00831 0.02653

6 5 0 00734 0.0257 0.00573 0.01822

7/ . nu 0 00527 0 01839 0.00394 0.01 249

7 5 0 00378 0 01 31 2 0.00271 0.00855

8o . nu 0 nn?70 0 00934 0 00186 0.00584

8.5 0.00193 0.00664 0.00127 0.00398

9.0 0.001 37 0.00471 0.00087 0.00271

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 0.0001 2 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);

^^^Spectrum fraction above lower energy bound

11
Spectrum uncertainties in the evaluation are based on standard deviations

from the mean of subsets of measured data from the final adjusted Maxwellian,

This estimate, carried out in a seven group structure, has been made for a la

uncertainty. Results are given in Table 4.

TABLE 4. UNCERTAINTY ESTIMATES FOR 252cf AND 235u EVALUATED FISSION NEUTRON

SPECTRA

Energy

Boundari es

252cf

(Spontaneous Fission)

Uncertai nty

235U

(Thermal -Neutron -Induced Fi ssion)

Uncertai nty

d I u u 1 n fir ni 1 n 1 a

Fl uence (%) Fl uence (%)

(}) • AE (j) • AE

0.047 ±13 0.054 ±16

0.184 ± 1.1 0.197 ± 4.1

0.220 ± 1.8 0.229 ± 3.0

0.194 ± 1.0 0.195 ± 3.1

0.200 ± 2.0 0.192 ± 2.0

0.146 ± 2.1 0.127 ± 4.8

0.0087 ± 8.5 0.0056 ± 5.3

(0.00058)* (0.00026)*

0.0

0.25

0.8

1.5

2.3

3.7

8

12

20

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

estimates.

III. DETAILED OPERATING PROCEDURES

Assembly and Irradiation of Sensors

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

on the plastic bag containing the sensor. Although the mass is not

necessary to specify the fluence, it is often useful for the calculation

of other parameters, e.g., scattering.

2) The sensors should be mounted in the lightweight holders ("spiders") and

subsequently in the mounts. Sensors are always mounted in pairs, on the

same axis and on opposite sides of the source. The positions of the sensors should be recorded on the worksheet and in the notebook.

12
3) Have the shops measure the separation distance of the sensor pairs on the

coordinate measuring machine. This should be a direct measurement of

face-to-face separation when possible. Record all parameters, so that a final center-to-center separation can be obtained. A notebook sketch of

the measurement procedure is advisable. The current shop contact is

Gene Morgan at extension 2448.

4) Arrange time at the Time-of -Fl i ght-Faci 1 i ty (TOFF) with Linac Operations,

CW Operations, and the Neutron T-O-F group. Notify Health Physics.

5) Mount the source and sensor holder ring at the TOFF and connect the

desired source to the beaded chain. For some irradiations, the cadmium

shield will be necessary and should be mounted at this time. Lock the

road gate with padlock and put up any required ropes and signs (requirements vary with source strength and length of irradiation). Be sure that

there are no persons at the TOFF end stations or in the blockhouse. Turn

on the red outside warning lights.

6) Check time marking instrument with telephone time (844-2525) or Naval

Observatory Time (653-0258). Raise the source while noting the time.

Note that the source up warning light has come on. After securing the

source, record the time in notebook and source activities log. Visually

inspect the source with telescope when possible.

7) On leaving, close and padlock locking gate 11; put up rope and sign in

hallway; close but do not lock door at top of stairwell; make sure that Cf source up sign is showing; leave sign with your name and phone numbers

showing; leave appropriate signs on external door and in Linac control

room.

8) At the end of the irradiation (EOI) , again check time. Lower the source

while noting the time. Be sure that the source up warning light goes out ! Record the EOI time in notebook and source activities log. Later be sure

that these times are recorded on the worksheets. Secure source, unlock

road gate, remove appropriate ropes and signs. Remove sensor holder.

9) Remove sensors from holders and package individually in labeled plastic

containers. Store in shielded card file in Building 235, Room B123.

IV. FLUENCE CALCULATIONS AND UNCERTAINTIES

IV. a. FLUENCE AND FLUENCE RATE CALCULATIONS

1) Obtain the source strength, S, from output of the FORTRAN program CFYLD

(See Appendix A). This program can be run for a specific day or can list

in tabular form the source strength for consecutive days. The source strength of the most recent manganous sulfate bath measurements of the source is entered and the program uses the 252cf hg-jf ]]fQ to calculate

the source strength on the days of interest.

13
2) The fluence rate, (j), averaged over the surface of the sensor is obtained

from the equation:

where R is the average source to sensor distance and p is the radius of

the sensor, typically 6.35 mm.

This equation has been programmed on a TI SR-52 desk calculator,

3) Run the program CFGFAC (See appendix B) to obtain the corrections for

decay within irradiation of both the Cf source and the activation

products of the sensors. This program will correct for interruptions

during the irradiation.

4) Since the scattering corrections differ for each reaction, scattering

corrections have been done on an ad hoc basis. Table 5 lists these

corrections for a few of the more common sensors.

5) Fill out test report (See appendix C). All reduced data should be kept

in the fluence standard file in room A156 of Building 235.

IV. b. UNCERTAINTIES

Table 1 gives uncertainty estimates for free-field fluence irradiations

at the Cf facility. The uncertainty in the distance measurement can be

considered a random uncertainty and part of the source strength uncertainty is

also random. The ^^^Cf source strength is determined by a ratio measurement

to NBS-I and therefore the uncertainty of the NBS-I source strength is

systematic and the scattering corrections are also systematic, while the

uncertainty of the ratio to NBS-I is random.

TABLE 5. NET SCATTERING CORRECTIONS FOR SOME COMMON SENSORS

Sensor ^c

ii5in(n,n') + 0.007 0.5

58Ni(n,p) - 0.001 0.5

5'^Fe(n,p) - 0.003 0.5

i03Rh(n,n'

)

+ 0.013 0.5

14
V. BENCHMARK REFERENCING

A benchmark neutron field is a wel 1 -character i 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 sensor to a known neutron fluence, xcf» ''^ ^ 252cf fission spectrum establishes a calibration factor,

Q

_ [observed sensor response]

IxQf neutron fluence] '

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

cross section, a^, for the sensor reaction to obtain a consistent benchmark

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

the spectrum averaged cross sections of the study field.

[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 counting system and known reaction

parameters, branching ratios, isotopic abundances, fission yields etc.), then

the "observed sensor response" in Eq. (1) becomes the reaction probability,

and G is an observed spectrum-averaged cross section for the fission

spectrum.] The calibration factor may then be applied to the investigation of

a study spectrum.

By irradiating samples identical to those irradiated in the 252cf fjeld,

the fluence of the study field, ^'g^jj^y. can be obtained as a simple ratio of sensor responses, D, and spectrum averaged cross sections, a, so that:

(2)

*^studv _ ^studv ^studv

D a $ X XX

In practice, [D^^^^y^D ) is simply the time history corrected counting

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

the Cf 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 is

that given in the test report and one then solves equation (2) for ^s^udy*

15
VI. SAFETY

The neutron sources used for these irradiation services range in source

strength from 1 to 5 x 10^ n/s. While these sources are not lethal,

mishandling can result in radiation exposures well over the legal limit for

radiation exposures. The sources should be handled remotely. If close

handling is necessary (e.g., when remounting the source on its holder), the

source should be placed in the water tank. New irradiation setups should be

first practiced with the dummy sources expressly designed for this purpose.

Personnel dosimeters must be worn at all times and a "chirper" is recommended

when close handling is necessary. Unattended irradiations require that the

area be locked to prevent accidental entry. Ropes with warning signs should

be placed at the 2.5 mR radiation level in accordance with NRC regulations.

Health Physics will assist in determining the rope locations if necessary.

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

(usually a few nanocuries). There is therefore no hazard from the samples,

but all radioactive shipping regulations must still be followed.

16
REFERENCES

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

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

[2] Grundl, J. A., Spiegel, V., Eisenhauer, C. M. , Heaton II, H. T.

,

Gilliam, D. M. NBS; and Bigelow, J., ORNL; "A Cal i f orni um-252 Fission

Spectrum Irradiation Facility for Neutron Reaction Rate Measurements," Nucl.

Tech. 32, 315 (flarch, 1977).

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

fleasurements in 252cf 2 3 5ij Fission Spectra," Proc. Int. Conf. on Nuclear Data for Science and Technology, Antwerp, Belgium (September, 1982).

[4] DeJuren, J. A., Padgett, D. W., and Curt^ss, L. F., J. Res. NBS 40, 63

( 1955)

.

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

[6] Noyce, R. H., Mosburg, E. R., J'^., Gar^finkel, 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) J7, 313 (1963).

[7] Spiegel, V., private communication.

[8] McGarry, E. D. and Boswell, E. W., "Neutron Source Strength by the Manganese

Sulfate Bath Method," NBSIR, to be published.

[9] Li Linpei, "Reflection of 252cf Fission Neutrons from a Concrete Floor,"

Radiation Protection Dosimetry, _5, 237 (1984).

[10] 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).

[11] Grundl, J. A. and Eisenhauer, C. M., "Fission Specti^um Neutrons for Cross

Section Validation and Neutron Flux Transfer," Proc. of Conf. on Nuclear

Cross Sections and Technology, NBS Special Publication 425, U.S. Dept. of

Comme:"ce, Washington, D.C. (March, 1975).

[12] 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).

[13] Madland, D. G., LaBauve, R. J., and Nix, J. R., " Di f f erenfi al and Integral

Comparisons of Three Representations of the Prompt Neutron Spectrum for the

Spontaneous Fission of 252Qf^ii Alamos National Lab. Report

LA-UR-84-3557 (1984).

[14] Wiedling, T., Proc. Symp. on Neutr^on 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).

17

APPENDIX A

C CFYLD--CALIFORNIUM YIELD 6/26/82.

DIMENSION KC0M(66)

CHARACTERAl REP

CHARACTERA 4 A

C THIS PROGRAM USES 2.645 YR HALFLIFE FOR 252CF AND NU=3.766.

C FOR 250CF IT USES 13.2 YRS AND 3.53. FOR 254CF IT USES 61.9 DAYS

C AND 3.93. NU RATIOS ARE FROM N. SC. AND ENGIN. 43,54(1971).

C RATIO OF ALPHA PARTICLE TO SPONTANEOUS FISSION RATE IS FROM

C J. INORG. NUCL CHEM 27,33(1965)., IE., 31.3 AND 1260.

C LAMBDA ALPHA/ (LAMBDA SP FISSION) = 1/(1260+1) AND 1/(31.3+1)

C SP FISSION/ALPHA FOR 254 IS 0.997

C

DATA CFOHF/13. 2/ , CF2HF/ 2 . 645/ , CF4HF/61.9/, CFONU/3.53/,

&CF2NU/3.766/ , CF4NU/3.93/

A= 'XXXX'

1 FORMAT(F14.0)

2 F0RMAT(6X,A4,I2, ' ,

' , 14 , 1P4E12 . 4

)

3 F0RMAT(I5)

4 FORMAT ( /7X,

S'GIVE THE DAY, MONTH, AND YEAR OF ISOTOPIC ANALYSIS ')

5 F0RMAT(7X, ' IN F FORMAT THE ATOM PERCENT COMPOSITION OF CF250,

& 252, AND 254 ='

)

6 F0RMAT(7X, 'GIVE THE DAY, MONTH, AND YEAR OF CALIBRATION ')

7 FORMAT (/7X' DATE ' 3X' 250 /TOTAL ' 3X' 254 /TOTAL ' 3X

&'DECAY'7X'N/SEC' /

)

8 FORMAT ( 7X' GIVE THE DAY, MONTH, AND YEAR FOR THE END OF THE ' &/7X' TABULATION '/)

9 F0RMAT(7X'IN F OR G FORMAT WHAT IS THE CALIBRATION

SON THAT DATE ? '

)

10 FORMAT ( 7X' GIVE THE DAY, MONTH, AND YEAR FOR THE START OF THE ' &/7X' TABULATION IN I FORMAT'/)

11 F0RMAT(/7X,

S'THE HALF LIVES USED IN YEARS FOR CF250 AND CF252,',/

&7X,' AND IN DAYS FOR CF254 = '/)

12 FORMAT (7X' TYPE THE TITLE INFORMATION '/)

13 F0RMAT(66A1)

14 F0RMAT(7X,66A1)

15 FORMAT( 7X,3F14.6)

16 FORMAT ( /7X, 'THE VALUES USED FOR NUBAR = '/)

18 FORMAT ( /7X' GIVE A CARRIAGE RETURN AFTER EACH DATA ENTRY')

19 FORMAT( 7X'THE ATOM PERCENT COMPOSITION OF CF250,

& 252 AND 254 =

'

)

20 FORMAT ( 7X ' WHENEVER YOU REQUIRE NO FURTHER OUTPUT, ANSWER THE NEXT

&/7X' REQUEST FOR DAY, MONTH, AND YEAR WITH THREE CARRIAGE RETURNS.

&' '/)

21 FORMAT ( /7X' THE OUTPUT WILL BE ON F0R002.DAT.')

22 FORMAT (Al)

23 FORMAT ( 7X' GIVE THE DAY, MONTH, AND YEAR '/)

24 F0RMAT(7X'D0 YOU WISH TO INPUT A SPECIFIC DATE (Y/N) ? '/)

OPEN ( 2 , STATUS = ' NEW ' ) WRITE(7,18)

WRITE(7,4)

READ(5,3)ID,M,IYR0

CALL IDAYS( ID,M,IYRO,IDAYO)

WRITE( 7,5)

A-1
READ( 5 , 1 )X050 ,X052 ,X054

C NORMALIZE THE NEUTRON EMMITTERS TO 1.

SN=X050+X052+X054

X50=X050/SN

X52=X052/SN

X54=X054/SN

WRITE(7,6)

READ( 5,3) ID,M,IYR1

CALL IDAYS(ID,M,IYR1,IDAY1)

CALL LPDAYS(IYR1,IYR0,ITL)

WRITE( 7,9)

READ(5,1)CA

IT=( lYRl-IYRO) A365+( IDAY1-IDAY0)+ITL

C * DEBUG OPTIONS FOLLOW

C WRITE( 7 , 400 ) IDAYO , lYRO , IDAYl , lYRl , ITL , IT

C 400 F0RMAT(7X, 'IDAYO, lYRO, IDAYl, lYRl, ITL, IT' ,/7X, 816)

T=FLOAT(IT)

A500=0.693147*CF0NUAX50/ (CF0HFA365.25A1261.

)

A520 = 0 . 693147 ACF2NU'^X52 /( CF2HFA365 . 25^^32 . 3 ) A540=0 . 693147ACF4NUA0 . 997AX54/CF4HF

AAT=A520+A500+A540

AT0=1.

F00=A500/AAT

F40=A540/AAT

A50=A500AEXP( -0. 69314 7AT/ (CF0HFA365. 25)

)

A52=A520AEXP (-0.693147 AT/ (CF2HFA365. 25)

)

A54=A540AEXP( -0 . 693147AT/CF4HF)

ATT=A50+A52+A54

AT1=ATT/AAT

F01=A50/ATT

F41=A54/ATT

AT=AT1

CN0=CA/AT1

WRITE( 7,12)

READ(5,13)KC0M

WRITE( 2,14)KC0M

WRITE(2,19)

WRITE( 2,15)X050,X052,X054

WRITE( 2,11)

WRITE (2,15) CFOHF, CF2HF, CF4HF

WRITE(2,16)

WRITE(2,15) CFONU, CF2NU, CF4NU

WRITE (2, 7)

CALL MODAY( IYRO,IDAYO,A,ID)

WRITE(2,2)A,ID,IYR0,F00,F40,AT0,CN0

CN1=CN0AAT1

CALL M0DAY(IYR1,IDAY1,A,ID)

WRITE(2,2)A',ID,IYR1 ,F01 ,F41,AT1,CN1

WRITE( 7,20)

60 WRITE(7,24)

ACCEPT 2 2, REP

IF ( (REP.EQ. ' Y' ) .OR. (REP.EQ. 'y ' ) ) GO TO 70

WRITE( 7,10)

READ(5,3)ID,M,IYR2

IF(ID.EQ.0)GO TO 200

CALL IDAYS( ID,M,IYR2,IDAY2)

WRITE( 7,8)

A-2

1
READ(5,3)ID,M,IYR3

CALL IDAYS(ID,M,IYR3,IDAY3)

CALL LPDAYS(IYR3,IYR2,ITL)

N=( IYR3-IYR2) A365+IDAY3-IDAY2+ITL+1

GO TO 80

70 N=l

WRITE( 7,23)

READ(5,3)ID,M,IYR2

IF(ID.EQ.0)G0 TO 200

CALL IDAYS(ID,M,IYR2,IDAY2)

80 CALL LPDAYS(IYR2,IYR0,ITL)

IT2= (IYR2-IYR0 ) A365+IDAY2-IDAY0+ITL

C A DEBUG OPTIONS FOLLOW

C WRITE( 7 r 500 ) IDAYO , lYRO , IDAYl , lYRl , IDAY2 , IYR2 , ITL , IT2

C 500 F0RMAT(7X, ' IDAYO , lYRO , IDAYl , lYRl , IDAY2 , IYR2 , ITL , IT2 ' ,/7X,8I6)

T=FLOAT{ IT2)

DO 100 1=1,

A50=A500AEXP( -0.693147 AT/ (CF0HFA365. 25)

)

A52=A520AEXP( -0.693147 AT/ (CF2HFA365. 25)

)

A54=A540AEXP( -0.693147AT/CF4HF)

ATT=A50+A52+A54

AT1=ATT/AAT

F01=A50/ATT

F41=A54/ATT

CN1=CN0AAT1

CALL M0DAY(IYR2,IDAY2,A,ID)

IF ( ( REP . EQ .' Y

' ) .OR . ( REP . EQ .

' y

' ) ) THEN

WRITE(7,2)A,ID,IYR2,F01,F41,AT1,CN1

END IF

WRITE(2,2)A,ID,IYR2,F01,F41,AT1,CN1

IDAY2=IDAY2+1

IF(IDAY2.LT.366)G0 TO 100

IF( IDAY2.GT. 366)G0 TO 95

IM0D=M0D(IYR2,4)

IF( IMOD.NE.0)GO TO 95

GO TO 100

95 IDAY2=1

IYR2=IYR2+1

100 T=T+1.

GO TO 60

200 CL0SE(2)

WRITE( 7,21)

STOP

END

C

C

SUBROUTINE IDAYS ( ID ,M , lYR , IDAY

)

C THIS SUBROUTINE CALCULATES THE JULIAN CALENDAR DAY OF THE YEAR

C

1 FORMAT ( 7X' YOU TYPED MONTH AND DAY. TYPE DAY AND MONTH '/)

3 FORMAT (15)

IF( lYR.LT. 1900) IYR=IYR+1900

IF(M.GT. 12)P^ITE( 7,1)

IF{M.GT. 12)READ( 5,3)ID,M

IY=MOD( IYR,4)

C FOR MONTHS AFTER FEBRUARY

X=2. 0

A3
1

I C FOR JANUARY AND FEBRUARY

IF(M.LE.2)X=0.0

C FOR MONTHS AFTER FEBRUARY IN LEAP YEARS

IF(M.GT.2.AND. IY.EQ.0)X=1.0

Y=(M-l)A30.57+0.5+ID-X ' IDAY=INT(Y)

RETURN

END

C : c

I SUBROUTINE LPDAYS ( I2YR , IIYR , ITL) ' C i C THIS SUBROUTINE CALCULATES THE LEAP YEAR DAYS BETWEEN II YR i C AND THE YEAR PRECEDING I2YR. 1 c i ITL=0 1 IF( IIYR.GE. I2YR)G0 TO 60 I DO 50 I=I1YR,I2YR-1 \ IM0D=M0D(I,4) i C ^ DEBUG OPTIONS FOLLOW | C 3 FORMAT ( 7X ITL , I , AND IMOD ARE ',318) ! C WRITE(7,3)ITL,I,IMOD j 50 IF( IMOD.EQ. 0) ITL=ITL+1 I 60 RETURN 1 END I C "i SUBROUTINE MODAY ( lYR , IDY , A, ID) 1 C i C MODAY- -THIS SUBROUTINE CALCULATES THE MONTH AND DAY OF THE YEAR ! C FROM THE JULIAN CALENDAR DAY. 9/11/81 i CHARACTERA4 AO ,AJAN,AFEB,AMAR,AAPR ,AMAY,AJUN,AJUL,AAUG,ASEP,AOCT, I I ANOV,ADEC,A i DATA AO/ 'AAAA' / ,AJAN/ ' JAN '/,AFEB/'FEB '/,AMAR/'MAR '/, \ &AAPR/'APR '/,AMAY/'MAY '/,AJUN/'JUN '/,AJUL/'JUL '/,AAUG/'AUG '/, j &ASEP/'SEP '/,AOCT/'OCT '/,ANOV/'NOV '/,ADEC/'DEC '/ I C i IF(IDY.EQ.O) GO TO 1000 | A=AO I J = l C

C IDY MUST NOT BE CHANGED BECAUSE IT IS RETURNED TO MAIN PGM

C

IDAY=IDY I IM0D=M0D(IYR,4) i IF(IDAY-31)10,10,20

10 A=AJAN

ID=IDAY

GO TO 1000

20 IF(IDAY-59)30,30,40 | 30 A=AFEB ! ID=IDAY-31 j GO TO 1000 i 40 IFdDAY. EQ. 60. AND. IMOD.EQ. 0)J = 0

] IF( J-l)50,60,60 j 50 A=AFEB I ID=IDAY-31 I AGO TO 1000

60 IF(IMOD.EO.O)IDAY=IDAY-1

C AAAAAAAAAA>;?kAAAAAAAAAAAAAAAAAAAAAAAAAAAAAAAAAAAAAAAAAAAAAA

C IDAY AT THIS LINE OF THE PROGRAM MUST BE 60 OR GREATER.

C THIS SECTION OF THE PROGRAM DECIDES WHERE THE PROGRAM

C SHOULD CONTINE IN ORDER TO AVOID MOST OF THE FOLLOWING 'IF'

C STATEMENTS

.

C

K=(IDAY-6) /30

GO TO (65,65,80,100,120,140,160,180,200,220,240,240) K

C

C AAAAAAAAAAAAAAAAAAAAAAAAAAAAAAAAAAAAAAAAAAAAAAAAAAAAAAAAAA

65 IF( IDAY-90)70,70,80

70 A=AMAR

ID=IDAY-59

GO TO 1000

80 IF(IDAY-120)90,90,100

90 A=AAPR

ID=IDAY-90

GO TO 1000

100 IF(IDAY-151)110,110,120

110 A=AMAY

ID=IDAY-120

GO TO 1000

120 IF(IDAY-181)130,130,140

130 A=AJUN

ID= IDAY- 151

GO TO 1000

140 IF( IDAY-212)150,150,160

150 A=AJUL

ID= IDAY- 181

GO TO 1000

160 IF(IDAY-243)170,170,180

170 A=AAUG

ID=IDAY-212

GO TO 1000

180 IF(IDAY-273)190,190,200

190 A=ASEP

ID=IDAY-243

GO TO 1000

200 IF(IDAY-304)210,210,220

210 A=AOCT

ID=IDAY-273

GO TO 1000

220 IF(IDAY-334)230,230,240

230 A=ANOV

ID=IDAY-304

GO TO 1000

240 A=ADEC

ID=IDAY-334

GO TO 1000

1000 RETURN

END

A5

APPENDIX B

A THIS VERSION OF CFGFAC IS DATED NOVEMBER 5,1985

DOUBLE PRECISION TMI ( 20 ) ,CMI ( 20

)

DOUBLE PRECISION TOUT ,DE ,HE ,TE ,DS ,HS ,TS ,TDI ,THI , CDI , CHI ,T

,

+FLU r PFLU r ALCF , TAU , ALAM , SATFR , B , GFAC , CFAC , BLAM

1 FORMATdS)

2 FORMAT(D16.0)

3 FORMAT (7X,

+/7X'THE G-FACTOR IN THIS PROGRAM IS THE SAME AS DEFINED IN'

&/7X'THE NBS COMPENDIUM AND THE C-FACTOR IS G/LAMBDA AND IS ' &/7X'DIMENSI0NLESS. IN THIS PROGRAM THE SOURCE IS 252CF WITH A'

+/7X' 2.645 YEAR HALF-LIFE.'/)

4 FORMAT( 7X'H0W MANY TIMES WAS THE IRRADIATION INTERRUPTED?')

5 FORMAT ( 7X' GIVE THE DAY, HOUR, AND MINUTE FOR THE END OF THE'

+/7X' IRRADIATION IN D FORMAT, ONE TO A LINE.'/)

6 FORMAT ( 7X' GIVE THE DAY, HOUR, AND MINUTE FOR THE START OF THE'

+' IRRADIATION'/)

7 FORMAT ( 7X' GIVE THE DAY, HOUR, AND MINUTE FOR THE FIRST ' +' INTERRUPTION' /

)

8 FORMAT ( 7X ' THE SAME FOR THE NEXT INTERRUPTION'/)

9 F0RMAT(7X'IN D FORMAT GIVE THE HALF-LIFE OF THE ACTIVATED ' + ' NUCLEI IN DAYS ' /

)

10 FORMAT ( 7X' TYPE A 1 TO INPUT ANOTHER HALF-LIFE'

)

11 FORMAT( 7X'THE G-FACTOR FOR THE ' , 1PD14.. 6 , +/7X'DAY HALF-LIFE IS ' , 1PD14 . 6

,

' PER SECOND'

&/7X' AND THE C-FACTOR IS ' , 1PD14 . 6

,

' AND IS DIMENSIONLESS '

)

13 FORMAT ( 7X' GIVE THE DAY, HOUR, AND MINUTE FOR THE CONTINUATION'

+/7X'0F THE IRRADIATION'/)

14 FORMAT ( 7X' THE TOTAL IRRADIATION TIME WAS', DIG. 8,' SECONDS'/)

15 FORMAT( 7X'THE IRRADIATION CYCLE LASTED ' ,D16 . 8

,

' SECONDS'/)

16 FORMAT (7X' THE TOTAL FLUENCE FOR A FLUX DENSITY OF 1 N/CMAA2-SEC'

&/7X'AT THE END OF THE IRRADIATION WAS ',1PD16.8,/)

18 F0RMAT(7X'IF THE CF SOURCE HAD A CONSTANT EMISSION RATE, THE'

&/7X' TOTAL FLUENCE FOR A FLUX DENSITY OF 1N/CMAA2-SEC AT THE START'

&/7X'0F THE IRRADIATION WOULD HAVE BEEN ',1PD16.8,/)

19 FORMAT (/7X' THE RESULTS ARE STORED ON F0R003.DAT.')

OPEN( 3, STATUS = 'NEW'

)

TOUT=0.D0

WRITE( 7,3)

WRITE( 7,4)

WRITE( 3,3)

READ(7,1)II

WRITE( 7,5)

A READ DAY, HOUR, AND MINUTE FOR THE END OF THE IRRADIATION.

READ(7,2)DE,HE,TE

WRITE(7,6)

A READ DAY, HOUR, AND MINUTE FOR THE START OF THE IRRADIATIION.

READ( 7 ,2)DS,HS,TS

TS=TS+ ( DSA24 . DO+HS ) A60 . DO

TE=TE+ ( DEA24 . DO+HE ) A60 . DO-TS

IF(II.Eg.O)GO TO 110

A READ THE INTERRUPTION AND RESTART TIMES

DO 100 1=1,11

IFd.EQ.DGO TO 101

WRITE(7,8)

B-1
READ( 7 , 2 )TDI ,THI ,TMI ( I

)

GO TO 102

101 WRITE(7,7)

READ( 7 ,2 )TDI ,THI ,TMI( I

)

102 WRITE(7,13)

READ ( 7 , 2 ) GDI , CHI , CMI (I

)

TMK I) =TMI{ I)+(TDIA24.D0+THI) A60.D0-TS

GMK I)=GMI( I)+(GDI^24.D0+GHI) A60.D0-TS

100 TOUT=GMI( I)-TMI( I)+TOUT

110 T=TE-TOUT

TSS=T'^60.D0

TES=TEA60.D0

WRITE ( 7,14) TSS

WRITE(7,15)TES

WRITE(3,14)TSS

WRITE( 3,15)TES

ALGF=-0.69314718D0/ ( 2 . 645D0A365 . 25D0A1440 . DO

)

PFLU=O.DO

BG=1.D0

IF(II.EQ.O)GO TO 180

DO 175 1=1,11

175 PFLU=PFLU+DEXP ( ALGFAGMI ( I )

) -DEXP ( ALGFATMI (I )

)

BC= ( 1 . DO-PFLU/ ( DEXP ( ALGFATE ) -1 . DO )

)

180 FLU=(BG/ALGF) A (1 .DO-DEXP( -ALGFATE) ) A60.D0

WRITE(3,16)FLU

VmiTEi 7 ,16)FLU

WRITE( 3,18)TSS

WRITE(7,18)TSS

200 WRITE(7,9)

READ(7,2)TAU

BLAM=0.69314718D0/ ( 1440.D0ATAU)

ALAM=BLAM+ALGF

PSATFR=0.D0

B=1.D0

IF(II.EQ.0)GO TO 260

DO 250 1=1,11

250 PSATFR=PSATFR+DEXP(ALAMAGMI ( I )

) -DEXP( ALAMATMI ( I )

)

B= ( 1 . DO-PSATFR / ( DEXP ( ALAMATE ) - 1 . DO )

)

260 SATFR=B/ALAMA( 1 .DO -DEXP( -ALAMATE) ) ABLAM

BLAM=BLAM/60.

GFAG=SATFR/FLU

GFAG=GFAC/BLAM

WRITE (7,11) TAU , GFAC , GFAG

WR ITE ( 3 , 1 1 ) TAU , GFAG , GFAG

WRITE( 7,10)

READ(7,1)MM

IF(MM.EQ. DGO TO 200

WRITE( 7,19)

GL0SE(3)

STOP

END

B-2
APPENDIX C

TEST REPORT FOR NEUTRON FLUENCE STANDARDS

PRODUCED BY AN IRRADIATION AT THE NBS ^^^Cf FISSION NEUTRON SOURCE

Irradiation No.: Report Date:

Neutron Source:

Sensor Materials Irradiated:

( designations

End of Irradiation (EOI):

REPORT FOR:

Free-field, fission-neutron fluence in compensated beam geometry (averaged

over the volume of the detectors)

4> = n/cm^

Physical description of material: Mass = Form: mm dia. X mm thk disk

Material:

Additional parameters of irradiation:

Decay correction factor: C = using X = Length of Irradiation: T = Departure from free-field activity due to neutron scattering:

The gradient of the neutron fluence across the thickness of the detector

package is %/mm. The fluence at the center of the detector disk is

% greater than at the edge.

Inventory of Uncertainties (1 o

)

(1) Neutron source strength of ^^^Cf 1.1%

(2) Distance measurement for compensated geometry 0.2%

(3) Neutron scattering including sensors 0.5%

Other Radioactivities

Since the sensors are natural metal samples, 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:

C-1
TEST REPORT FOR NEUTRON FLUENCE RATE STANDARDS

PRODUCED BY AN IRRADIATION AT THE NBS ^^^Cf FISSION NEUTRON SOURCE

Irradiation No.: Report Date

Neutron Source:

Sensor Materials Irradiated:

( designations

End of Irradiation (EOI):

REPORT FOR:

Free-field, fission-neutron fluence rate in compensated beam geometry

averaged over the volume of the detectors)

<p = n/cm^s

Physical description of material: Mass =

Form: mm dia. X mm thk disk

Material

Additional parameters of irradiation:

Decay correction factor: [ATC] = using X = Length of irradiation: T = Departure from free-field activity due to neutron scattering:

The gradient of the neutron fluence rate across the thickness of the

detector is %/mm. The fluence at the center of the detector disk is

% greater than at the edge.

Inventory of Uncertainties (la)

(1) Neutron source strength of ^^^Cf ^.^%

(2) Distance measurement for compensated beam geometry 0.2%

(3) Neutron scattering including sensors 0.5%

Other Radioactivities:

Since the sensors are natural metal samples, 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:

C-2
Comments

1) This service is offered as SP No. 44080C. A document describing the

service is in preparation and will be forwarded when available.

2) Reference Docoument: "Compendium of Benchmark Neutron Fields for

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

This fluence standard prepared by:

Neutron Dosimetry Group

Reviewed by:

J. A. Grundl

Neutron Dosimetry Group

For the Director

R. S. Caswell, Chief

Ionizing Radiation Division

Distribution (File copy only)

laboratory contact

test

report

sent

fluence

standard

sent
Relationship of Test Report Data to the Measured Activity at End of

252 Irradiation (EOI) for Cf Fission Neutron Irradiations

Compensated beam geometry

Neutron fluence standards are irradiated in

pairs on opposite sides and equidistant from the source. In this

arrangement, which is called compensated beam geometry, the fluence or fluence rate given in the test report is related to the geometric mean,

A , of the two measured detector activities:

m

A = /A, A, = [const.] x [fluence or fluence rate]. (1) m ' ^

Average neutron fluence or fluence rate

This test report certifies a neutron fluence or fluence rate which is a free-field value (i.e. neutron

scattering is neglected) averaged over the volume of the detectors.

A correction for neutron scatteringCy^^) may be included in the test

report. Based on a neutron transport calculation, it is equal to the

fractional departure from free-field activation attributable to neutron

scattering

A = A (1 + y ) (la) m sc

where A = mean of detector activities measured at EOI (Eq. 1 ) _m

A = net free-field mean of detector activities at EOI

M = 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 activity

of the neutron fluence standard in disintegrations per second is related

to the free-field fission neutron fluence by the activation equation:

= 0 • $ (2) XCN

where: A = net free-field mean of detector activities at EOI (dps)

from Eq. la

X = decay constant of reaction product

C-4
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( -AT ) ) /AT] . (When it is

appropriate to report average neutron fluence rate, the decay

correction factor is given as the product ATC — see below.)

N = number of reaction isotope atoms in the fluence standard

2

0 = fission-spectrum-averaged reaction cross section (cm ) 2

O = free-field fission neutron fluence (n/cm ). (Note: the symbol

<I> is recommended by ICRU and ASTM El 70.)

0'$ = reaction probability

Activation equation f or 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): — = a ' <<t>> (3)

[ATC]N

= saturated specific activity

[ATC]N

[ATC] = decay correction factor for average neutron fluence rate. For

an uninterrupted irradiation of lengt^i^T at constant

fluence rate, [ATC] is equal to [1-e ].

T = length of irradiation given in the test report.

<<|)> = the neutron fluence rate (n/cm^s)

a<(()> = 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'$) , (M)

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- (o-$) , (5)

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

daughter.

C-5
Recommended Procedure for Establishing a Calibration Factor (x^^)

Use Equations (2) or (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

detector cross section set as is used to obtain spectrum-averaged detector

cross-sections for subsequent neutron field measurements.

Revised: November 3, 1986

C-6
NBS-114A iREv. 2.BC) U.S. DEPT. OF COMM.

BIBLIOGRAPHIC DATA

SHEET (See instructions)

1. PUBLICATION OR

REPORT NO.

NBS/SP-250/13

2. Performing Organ. Report No. 3. Publ icatlon Date March 1988

4. TITLE AND SUBTITLE

NBS Measurement Services: Activation Foil Irradiation with

Californium 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 88-600508

I I Document describes a computer program; SF-185, PIPS Software Summary, is attached.

11. ABSTRACT (A 200-word or less factual sumnr]ary of most significant information,

bibliography or literature survey, mention it here)

if document includes a si gnifi cant

This document describes the NBS calibration service 4408nc

(formerly 8.10), 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 certified neutron fluence in a 252Qf spontaneous fission 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 1 x 10^^

neutrons/cm2 and maximum fluence rates of 1 x lO^ neutrons/cm^ sec,

with combined (la) uncertainties of ± 1,2%.

12. KEY WORDS (Six to twelve entries; alphabetical order; capitalize only proper names; and separate key words by semicolon s) activation; californium; dosimetry; fast neutrons; fission spectrum;

neutron fluence; neutron fluence rate

13. AVAILABILITY

Unlimited

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

38

15. Price

USCOMM-DC 6043-P80

1]IIiI jIiijI!I!I i

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 the 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 Mpha-Partide Calibrations SN003-003-0283-1 $2.00

SP 250-6 Regular Spectra! Transmtttance 003-003-02805-3 $3.25

SP 250-7 Radiance Temperature Calibrations 003-003-02827-H4 $2.25

SP 250-8 Spectral Reflectance 003-003-02812-6 $7.00

SP 250-9 Calibration of Beta-Partide-Emitting

Ophthalmic Applicators

003-003-02817-7 $2.00

SP 250-10 Radioactivity Calibrations with the "4ir" Gamma Ionization Chamber, and Other

Radioactivity Calibration Capabilities

003-003-02824-0 $2.25

SP 250-1 1 Dosimetry for Hig^Dose Applications

SP 250-12 Neutron Personnel Doslmetiy

003-003-02811-8 $2.50

SP 250-13 Activation Fofl 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 Instmments

SP 250-17 The NBS Photodetector Spectral Response

Calibration Transfer F*rogram

SP 250-18 Neutron Source Strength Calibrations

SP 250-19 Calibration of Gamma-Ray-Emitting

Brachytherapy Sources

SP 250-20 Spectral inadiance Calibrations 003-003-0282^1 $5.50

SP 250-21 Calibration of Beta-Particie Radiation Instrumentation

SP 250-22 Platinum Resistance Thermometer

Calibrations 003-003-02831-2 $17.00

SP 250-23 Uquid-in-GlassThemiometer

Calitiration Sen/ice

SP 250-24 Standard Cell Calibrations 003-003-02825-8 $2.75

SP 250-25 Calibration Sen/Ice for Inductive Voltage Dividers

SP 250-26 Phase Angle Calibrations

SP 250-27 AC-DC Difference Calibrations

SP 250-28 Solid-state DC Voltage 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

* Those entries containing a stock number (003-003

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