TABLE 3. EVALUATED FISSION NEUTRON SPECTRA FOR 252cf AND 235uU
IN 45-GROUP FORMAT

252¢f 235y
Lower «(a) « (a)
Energy Group Cumulative Group Cumulative
Boundary Fluence Fluence to Lower Fluence Fluence to Lower

(MeV) o ° AE Boundary o ° AE 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

0.2 0.0140 0.9668 0.0154

0.25 0.0152 0.9528 0.0166

0.3 0.0323 0.9377 0.0350

0.4 0.0337 0.9054 0.0363

0.5 0.0343 0.8717 0.0367

0.6 0.0343 0.8374 0.0365

0.7 0.0338 0.8031 0.0358

0.8 0.0664 0.7693 0.0697

1.0 0.0641 0.7029 0.0668

1.2 0.0608 0.6388 0.0629

1.4 0.0290 0.5780 0.0298

1.5 0.0279 0.5490 0.0286

1.6 0.0527 0.5211 0.0536

1.8 0.0484 0.4684 0.0487

2.0 0.0442 0.4200 0.0441

2.2 0.0206 0.3758 0.0204

2.3 0.0196 0.3552 0.0193

2.4 0.0364 0.3356 0.0356

2.6 0.0328 0.2992 0.0319

2.8 0.0296 0.2664 0.0284

3.0 0.0503 0.2368 0.0477 0.2095

3.4 0.0310 0.1865 0.0290 0.1618

3.7 0.0413 0.1555 0.0378 0.1328

4.2 0.0253 0.1142 0.0227 0.09503

4.6 0.0200 0.0889 0.0176 0.07233

5.0 0.0190 0.0689 0.0164 0.05473

5.5 0.0140 0.0499 0.0118 0.03833

6.0 0.0102 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 0.00137 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
i 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 10 MeV, use the shape function E% + exp(-aE);
a= 0.70 for 252cf; a = 0.76 for 235uU,
(a)

Spectrum fraction above lower energy bound.

ll


Relationship of Test Report Data to the Measured Activity at End of
Irradiation (EOI) for 252

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:

a = WVA,A, = Cconst.] x [fluence or fluence rate]. (1)

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 ecattering(u.,) 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:

aA, =A + ug.) (1a)

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

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

Uso = fractional departure from free-field activity due to
neutron scattering. (A minus sign means that the observed
activity 1s 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:

A
wee a ge 2
Ton =o °? (2)
where: A = net free-field mean of detector activities at EOI (dps)
from Eq. la
d = decay constant of reaction product

c-4


The californium source capsule, shown in figure 3, is made up of a disk-
shaped 252Cf 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 platinum-
iridium 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% (10). 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% (lo).

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

Free-field fission neutron fluence rate

(5-cm distance; source 6 x 109 n/s) 2 x 107 n/(cm2 s)
Source decay rate 2.2% / month
Typical maximum fluence 1013 n/cm2
Source capsule and support scattering correction < 1.3%

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

* Gamma-ray exposure at 5 cm ~ 170 R/h
(2.8 yr after separation)

Uncertainties (1c) 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


V. BENCHMARK REFERENCING

A benchmark neutron field is a well-characterized 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, xcr, in a
252Cf fission spectrum establishes a calibration factor,

G = [observed sensor response] (1)
[xp neutron fluence]

which may be used in conjunction with the calculated fission-spectrum-averaged
cross section, By5 for the sensor reaction to obtain a consistent benchmark
calibration. The cross section, o,, should be calculated with (1) the fission
spectrum shape associated with the%a priori spectrum (e.g., the source spec-
trum 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 field,
the fluence of the study field, Study? Can be obtained as a simple ratio of
sensor responses, D, and spectrum averaged cross sections, co, so that:

D study _ study ® study (2)
D 3. t)
x x x
In practice, Oceuay/?,) 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 9, is

that given in the test report and one then solves equation (2) for ®study°

15


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.
Grund], 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 Chris E. Kuyatt
Acting Chief Director
Measurement Services Center for Radiation Research

iii


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):
o= n/em’*s
Physical description of material: Mass =

Form: mm dia. X mm thk disk
Material:

Additional parameters of irradiation:
Decay correction factor: [ATC] = using \ =
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 (10)

(1) Neutron source strength of ?5?cf 1.1%
(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:


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 lo
uncertainty. Results are given in Table 4,

TABLE 4, UNCERTAINTY ESTIMATES FOR 252Cf AND 235U EVALUATED FISSION NEUTRON

SPECTRA
252CF 235y
(Spontaneous Fission) (Thermal -Neutron-Induced Fission)
Uncertainty Uncertainty
Energy Group lo Group lo
Boundaries Fluence (%) Fluence (4%)

> ° AE > * AE

ono 0.047 13 0.054 16
. + . +

0.25
0.8 0.184 #1.1 0.197 + 4.1
Ls 0.220 + 1.8 0.229 + 3.0
03 0.194 + 1.0 0.195 £3.1
37 0.200 + 2.0 0.192 + 2.0
3 0.146 + 2.1 0.127 + 4.8

0.0087 + 8.5 0.0056 + 5.3
12 * *
20 (0.00058) (0.00026)

*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


aqaaqagqaaaan

APPENDIX A

CFYLD--CALIFORNIUM YIELD 6/26/82.

DIMENSION KCOM(66)

CHARACTER*1 REP

CHARACTER*4 A

THIS PROGRAM USES 2.645 YR HALFLIFE FOR 252CF AND NU=3.766.
FOR 250CF IT USES 13.2 YRS AND 3.53. FOR 254CF IT USES 61.9 DAYS
AND 3.93. NU RATIOS ARE FROM N. SC. AND ENGIN. 43,54(1971).
RATIO OF ALPHA PARTICLE TO SPONTANEOUS FISSION RATE IS FROM
gd. INORG. NUCL CHEM 27,33(1965)., IE., 31.3 AND 1260.
LAMBDA ALPHA/(LAMBDA SP FISSION) = 1/(1260+1) AND 1/(31.3+1)
SP FISSION/ALPHA FOR 254 IS 0.997

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 FORMAT(6X,A4,12,',',14,1P4E12.4)
3 FORMAT(I5)
4 FORMAT(/7X,
&'GIVE THE DAY,MONTH, AND YEAR OF ISOTOPIC ANALYSIS ‘)
5 FORMAT(7X,‘IN F FORMAT THE ATOM PERCENT COMPOSITION OF CF250,
& 252, AND 254 =")
6 FORMAT(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 FORMAT(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 FORMAT(/7X,

&‘THE HALF LIVES USED IN YEARS FOR CF250 AND CF252,',/
&7X,' AND IN DAYS FOR CF254 = '/)

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

13 FORMAT(66A1)

14 FORMAT(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 FOR002.DAT.’)

22 FORMAT(A1)

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

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

OPEN( 2, STATUS= ‘NEW’ )
WRITE(7,18)

WRITE(7,4)
READ(5,3)1D,M,IYRO

CALL IDAYS(ID,M,IYRO,IDAYO)
WRITE(7,5)

A-1


5)

IV.b.

at th
cons
also
to NB
syste
uncer’

The fluence rate, ¢, averaged over the surface of the sensor is obtained
from the equation:

S R2
6 = gigs (RE tnli+52/R2]}

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.

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.

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.

Fill out test report (See appendix C). All reduced data should be kept
in the fluence standard file in room A156 of Building 235.

UNCERTAINTIES

Table 1 gives uncertainty estimates for free-field: fluence irradiations
e Cf facility. The uncertainty in the distance measurement can be

dered a random uncertainty and part of the source strength uncertainty is
random, The 252Cf source strength is determined by a ratio measurement
S-I and therefore the uncertainty of the NBS-I source strength is

matic and the scattering corrections are also systematic, while the
‘tainty of the ratio to NBS-I is random.

TABLE 5, NET SCATTERING CORRECTIONS FOR SOME COMMON SENSORS

Sensor Use bug¢(%) (10)
115In(n,n') + 0.007 0.5
58Ni(n,p) - 0.001 0.5
54Fe(n,p) - 0.003 0.5
103Rh(n,n') + 0.013 0.5

14


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)

test fluence
report standard
laboratory contact sent sent



where p = fraction of the neutron spectra above E
group display, $(E)AE, as follows:

p and in a coarse seven-

E limits: 0) 0.25 0.8 1.5 2.3 3.7 8 12 MeV
o(E)dE (252cCf) 0.047 0.184 0.220 0.194 0.200 0.146 0.009
o(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.

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

The neutron emission rate for the singly-encapsulated sources employed at
NBS are ~ 6 x 109 n/s when fabricated. The sources are periodically
calibrated in a MnSO, 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 107 n/(cm2 s) at the 5-cm distance, the fluence for a
one week exposure is ~ 1 x 1013 n/cm2, Spatial gradients for uncollided
fission neutrons are just those of a point source in free space:
(Ao/¢) = (2sr/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,

_ S
<P = G(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:

/D Dy co
= ao <6
Ny Ny

where D/N = sensor response per atom

o = 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,/N, within 20% of D,/N,.]


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 1013 neutrons/cm2 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 252Cf 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+CeNeoes (1)
where: A = free-field activity at EOI (8q),

A = decay constant of the reaction product (s~!),

C = decay correction factor (dimensionless) given in the test report
(this corrects for activity lost during the irradiation),

N = number of atoms of the target isotope,

a
u

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

and 4 = free-field fission neutron fluence (n/cm2) given in the test
report.

The commonly employed reaction rate probability (experimentally, the saturated
specific activity) is given by

A__-
xcTN = 7° ¢
where: sch = reaction rate probability (disintegrations per second per
nucleus),
= (@/T), the effective neutron fluence rate (n/cm2 s) or flux
density, and
T = the length of the irradiation given in the test report.


READ(7,2)TDI,THI,TMI(1I)
GO TO 102
101 WRITE(7,7)
READ(7,2)TDI,THI,TMI(I)
102 WRITE(7,13)
READ(7,2)CDI,CHI,CMI(I)
TMI (I) =TMI(1)+(TDIA24.DO+THI)*60.D0-TS
CMI(1)=CMI(1I)+(CDIA24.D0+CHI)*60.D0-TS
100 TOUT=CMI(1I)-TMI(I)+TOUT
110 T=TE-TOUT
TSS=TA60.DO
TES=TE*60.DO
WRITE(7,14)TSS
WRITE(7,15)TES
WRITE(3,14)TSS
WRITE(3,15)TES
ALCF=-0.69314718D0/(2.645D0*365.25D041440.D0)
PFLU=0.D0
BC=1.D0
IF(II.EQ.0)GO TO 180
DO 175 I=1,I1I
175 PFLU=PFLU+DEXP(ALCFACMI (I) )-DEXP(ALCF*TMI (TI) )
BC=(1.D0-PFLU/ (DEXP(ALCFATE)-1.D0) )
180 FLU=(BC/ALCF)%(1.D0-DEXP(-ALCFATE) )*60.D0
WRITE(3,16)FLU
WRITE(7,16)FLU
WRITE(3,18)TSS
WRITE(7,18)TSS
200 WRITE(7,9)
READ(7,2)TAU
BLAM=0 .69314718D0/(1440.DOATAU)
ALAM=BLAM+ALCF
PSATFR=0.D0
B=1.D0
IF(II.EQ.0)GO TO 260
DO 250 I=1,II
250 PSATFR=PSATFR+DEXP ( ALAMACMI (1) )-DEXP(ALAMATMI(T) )
B=(1.D0-PSATFR/ (DEXP(ALAMATE)-1.D0) )
260 SATFR=B/ALAMA(1.D0-DEXP( -ALAMATE) ) ABLAM
BLAM=BLAM/60.
GFAC=SATFR/FLU
CFAC=GFAC/BLAM
WRITE(7,11)TAU,GFAC ,CFAC
WRITE(3,11)TAU,GFAC ,CFAC
WRITE(7,10)
READ(7,1)MM
IF(MM.EQ.1)GO TO 200
WRITE(7,19)
CLOSE(3)
STOP
END


aaa aqaaqaaq aa

aaqaaaq a

aaa

*

FOR JANUARY AND FEBRUARY
IF(M.LE.2)X=0.0

FOR MONTHS AFTER FEBRUARY IN LEAP YEARS
IF(M.GT.2.AND.1Y.EQ.0)X=1.0
Y¥=(M-1)430.57+0.5+ID~-X

IDAY=INT(Y)

RETURN

END

SUBROUTINE LPDAYS(1I2YR,11YR,ITL)

THIS SUBROUTINE CALCULATES THE LEAP YEAR DAYS BETWEEN I1YR
AND THE YEAR PRECEDING I2YR.

ITL=0

IF(I1YR.GE.I2YR)GO TO 60

DO 50 I=I1YR,I2YR-1

IMOD=MOD(I,4)
DEBUG OPTIONS FOLLOW

3 FORMAT(7X,‘ITL,I, AND IMOD ARE ',318)
WRITE(7,3)ITL,I,IMOD

50 IF(IMOD.EQ.0)ITL=ITL+1
60 RETURN

10

20
30

40

50

END
SUBROUTINE MODAY(IYR,IDY,A,ID)

MODAY--THIS SUBROUTINE CALCULATES THE MONTH AND DAY OF THE YEAR
FROM THE JULIAN CALENDAR DAY. 9/11/81

CHARACTER*4 AO,AJAN,AFEB,AMAR,AAPR,AMAY,AJUN,AJUL,AAUG,ASEP ,AOCT,
1 ANOV,ADEC ,A

DATA AO/'x*xAK'/,AJAN/‘JAN ‘/,AFEB/‘FEB ‘/,AMAR/‘MAR ‘/,
&AAPR/’APR '/,AMAY/’MAY ’/,AJUN/’JUN ‘/,AJUL/’JUL ’/,AAUG/’AUG ‘/,
&ASEP/‘SEP ‘/,AOCT/‘OCT ‘/,ANOV/‘NOV ‘/,ADEC/’DEC ‘/

IF(IDY.EQ.0) GO TO 1000
A=A0
J=1

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

IDAY=IDY
IMOD=MOD(IYR,4)
IF(IDAY-31)10,10,20
A=AJAN

ID=IDAY

GO TO 1000

IF( IDAY-59)30,30,40
A=AFEB

ID=IDAY-31

Go TO 1000

IF (IDAY.EQ.60.AND.IMOD.EQ.0)J=0
IF(J-1)50,60,60
A=AFEB

ID=IDAY-31


TABLE OF CONTENTS

I, DESCRIPTION OF SERVICES AND FACILITIES

I.a. Generic Description. .....

I.b. Available Fission Neutron Exposures . .
I.c. Corrections for Neutron Scattering at 5
I.d. Configuration and Characteristics...

Il. NEUTRON FIELD CHARACTERIZATION ...
II.a. Neutron Fluence ......s
IIl.b. Neutron Spectrum... 6 ewe

II.b.1. Measurement .....

Il.b.2. Evaluated Spectrum 2... 2... 2.22222 2020-8

III. DETAILED OPERATING PROCEDURES...

IV. FLUENCE CALCULATION AND UNCERTAINTIES

IV.a. Fluence and Fluence Rate Calculations

IV.b. Uncertainties ......-e-

V. BENCHMARK REFERENCING... ee woe

VI, SAFETY 2 ew ee ee we ee we eee
REFERENCES 2 2 we ee ee eee we eee

APPENDIX A. Listing of CFYLD ......

APPENDIX B. Listing of CFGFAC......

APPENDIX C. Sample Test Reports ....

cm Position

NAY NHAY

fo)

10
10
10
12
13
13
14
15

16
7

A-1

B-1

C-1


ABSTRACT

This document describes the NBS calibration service 44080C
(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 252Cf 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 1013
neutrons/cm2 and maximum fluence rates of 1 x 107 neutrons/cm2 sec,
with combined (lo) uncertainties of + 1.2%.

Key words: activation; californium; dosimetry; fast neutrons; fission
spectrum; neutron fluence; neutron fluence rate


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

II.b.1. 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, well-
documented measurements of 252Cf and 235 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 252Cf and 235U fission spectrum shapes recommended in the evalua-
tion fg 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

piecewise-continuous 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 VE + exp(-1.5€/2.13) , E in MeV,
with adjustment functions u(E) as follows:

Energy Interval (MeV) Yor (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
1

+

1.5 - 6.0 - 0.0006E + 0.0037
6.0 -20 0 expl-0.03(E-6.0)]

The evaluated spectrum then is given by x(E) = u(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 235U evalu-
ated spectrum for comparison purposes, For other energy group structures this
tabulation may be interpolated as recommended in the footnote of Table 3.

10


LIST OF FIGURES

Page

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

Fig. 2 Typical arrangement of six activation sensors around the
252Cf source 2 ee eee ee ee ee ee ee ee ee eee ee 4
Fig. 3. Scattering materials in the low-mass 252Cf fission neutron
source capsule

ee

vii


VI. SAFETY

The neutron sources used for these irradiation services range in source
strength from 1 to 5 x 109 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


NDS MedSureiient OFIVILeS.

Activation Foil
Irradiation with
Californium Fission Sources

NBS
Special
Publication
250-13 .

George P. Lamaze
James A. Grund!

U.S. Department of Commerce
National Bureau of Standards

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I.a. GENERIC DESCRIPTION

A standard 252Cf neutron field consists of neutrons from the spontaneous
fission of 252Cf with 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 107 n/(cm2 s) are
obtained in isolated environments near small, intense 252Cf 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
1013 n/cm2 may be obtained with uncertainties as low as + 1.2% (lo).

Measurements of the 252Cf fission neutron spectrum and the closely
related the 235i) fission neutron spectrum are extensive and well-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- spect rum- -like components.

I.b. AVAILABLE FISSION NEUTRON EXPOSURES

e fluence rate (5 cm from source) ~ 2 x 107 n/(cm?2 s)
e nominal maximum fluence 1 x 1013 n/cm2
e accuracy of free-field fluence (1c) + 1.2%

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

e@ Net perturbation for threshold sensor reaction rates: < (1.0 + 0.8)%
(1c)

© Net perturbation for 235U (n,f) sensor reaction rate: (1.2 + 0.4)%
(10)

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.


APPENDIX B

* 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,PFLU,ALCF,TAU,ALAM,SATFR,B,GFAC,CFAC, BLAM
1 FORMAT(I5)
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'DIMENSIONLESS. IN THIS PROGRAM THE SOURCE IS 252CF WITH A’
+/7X'2.645 YEAR HALF-LIFE. ‘/)
4 FORMAT(7X‘HOW 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 FORMAT(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'OF THE IRRADIATION’ /)

14 FORMAT(7X‘THE TOTAL IRRADIATION TIME WAS‘ ,D16.8,'’ SECONDS’/)

15 FORMAT(7X‘THE IRRADIATION CYCLE LASTED‘ ,D16.8,‘ SECONDS’/)

16 FORMAT(7X‘THE TOTAL FLUENCE FOR A FLUX DENSITY OF 1 N/CM&A2-SEC’
&/7X'AT THE END OF THE IRRADIATION WAS ',1PD16.8,/)

18 FORMAT(7X‘IF THE CF SOURCE HAD A CONSTANT EMISSION RATE, THE’
&/7X'TOTAL FLUENCE FOR A FLUX DENSITY OF 1N/CM&A2-SEC AT THE START’
&/7X'OF THE IRRADIATION WOULD HAVE BEEN ',1PD16.8,/)

19 FORMAT (/7X' THE RESULTS ARE STORED ON FORO03.DAT.’)

OPEN (3 ,STATUS=’ NEW’ )
TOUT=0.DO
WRITE(7,3)
WRITE(7,4)
WRITE(3,3)
READ(7,1)1I
WRITE(7,5)
* READ DAY, HOUR, AND MINUTE FOR THE END OF THE IRRADIATION.
READ(7,2)DE,HE,TE
WRITE(7,6)
* READ DAY, HOUR, AND MINUTE FOR THE START OF THE IRRADIATIION.
READ(7,2)DS,HS,TS
TS=TS+(DS&24.D0+HS)*60.D0
TE=TE+(DE*24.D0+HE)460.D0-TS
IF(II.EQ.0)GO TO 110
* READ THE INTERRUPTION AND RESTART TIMES
DO 100 I=1,II
IF(I.EQ.1)GO TO 101
WRITE(7,8)


aaqaaaaa

ana

Go TO 1000
60 IF(IMOD.EQ.0) IDAY=IDAY-1
KRAAKKAKAAAKAAKAAAAKKAAKAAAAAKAKARAKAKAAAAKAAKAAAAAKKAARAAAARKAAA AKA
IDAY AT THIS LINE OF THE PROGRAM MUST BE 60 OR GREATER.
THIS SECTION OF THE PROGRAM DECIDES WHERE THE PROGRAM
SHOULD CONTINE IN ORDER TO AVOID MOST OF THE FOLLOWING ‘IF’
STATEMENTS.

K=(IDAY-6)/30
GO TO (65,65,80,100,120,140,160,180,200,220,240,240) K

AARKAAKKARAKKKAKAAKAKAKAAAAKAAKKAKKAKARAAAAAKAKAKAKARKAAKAAAAKRKARKRK
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 TFC 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


LIST OF TABLES

Page
TABLE 1. 252Cf Fission Neutron Field Parameters and Uncertainty
Components 2... ee ee ee eee we ee wee wee ee we OF

TABLE 2. Typical Neutron Fluence and Sensor Response Perturbations
due to Neutron Scattering... 1. eee eee eer ee eee 8B

TABLE 3. Evaluated Fission Neutron Spectra for 252¢f and 235U in
45-Group Format . 2. we ee ee eee ee eee eee ee ee ll

TABLE 4. Uncertainty Estimates for 252Cf and 235U Evaluated Fission
Neutron Spectra . 2 ew ee ee ee ee ee ee ee ee ene

TABLE 5. Net Scattering Corrections for some Common Sensors ...... 14

vi


3)

9)

IV.

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,

Arrange time at the Time-of-Flight-Facility (TOFF) with Linac Operations,
CW Operations, and the Neutron T-0-F group. Notify Health Physics.

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 (require-
ments 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.

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.

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.

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.

Remove sensors from holders and package individually in labeled plastic
containers. Store in shielded card file in Building 235, Room B123.

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 half life to calculate
the source strength on the days of interest.

13


Recommended Procedure for Establishing a Calibration Factor Oop?

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


READ(5,1)X050,X052,X054
c NORMALIZE THE NEUTRON EMMITTERS TO 1.
SN=X050+K052+X054
X50=X050/SN
X%52=X052/SN
X54=xX054/SN
WRITE(7,6)
READ(5,3)ID,M,IYR1
CALL IDAYS(ID,M,IYR1,IDAY1)
CALL LPDAYS(IYR1,IYRO,ITL)

WRITE(7,9)
READ(5,1)CA ;
IT=( IYR1-IYRO)A365+( IDAY1-IDAYO)+ITL

C *. DEBUG OPTIONS FOLLOW

c WRITE(7,400) IDAYO,1YRO,IDAY1,1YR1,ITL,IT

C 400 FORMAT(7X,’IDAYO,IYRO,IDAY1,IYR1,ITL,IT’ ,/7X,816)
T=FLOAT (IT)

A500=0.693147ACFONUAX50/ ( CFOHFA365. 2541261.)
A520=0.693147ACF2NUAX52/ (CF2HF*365. 2532.3)
A540=0.693147ACF4NUA0. 9974X54/CF4HF
520+A500+A540

A50=A500AEXP(-0.693147AT/ (CFOHF*365.25) )
A52=A520XEXP(-0.693147AT/ (CF2HF*365. 25) )

F41=A54/ATT

AT=AT1
CNO=CA/AT1
WRITE(7,12)
READ(5,13)KCOM
WRITE(2,14)KCOM
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,IDAY0O,A,ID)
WRITE(2,2)A,ID,IYRO,F00,F40,AT0,CNO
CN1=CNOAAT1
CALL MODAY(IYR1,IDAY1,A,ID)
WRITE(2,2)A,ID,IYR1,F01,F41,AT1,CN1
WRITE(7,20)

60 WRITE(7,24)
ACCEPT 22,REP

IF ( (REP.EQ.’Y’).OR.(REP.EQ.’y') ) GO TO 70
WRITE(7,10)
READ(5,3)1ID,M,IYR2

IF(ID.EQ.0)GO TO 200
CALL IDAYS(ID,M,IYR2,IDAY2)
WRITE(7,8)

A-2


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—materlals and instruments —are identified in this 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. Government Printing Office, Washington, DC 20402-9325


SOURCE
GUIDE

L~ TUBE

STAINLESS STEEL

Joes LID (0.59q)

Cf DEPOSIT (~3mg)

ecccc

7.9mm H i
wee ed

0.50 mm

(Cty 0280, DISPLACEMENT
~1.4 mm>)

ALUMINUM PELLET
(6.4 mm dia. x 5.7mm

Pa LONG ; 0.489)

STAINLESS STEEL CAPSULE
Pa (TYPE 304; 7.6mm dia,
0.809)

———— 76mm ————

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


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
source capsule scattering 0
support structure scattering 0
air scatter <0.

Net perturbation of sensor response due to neutron scattering:*
235y(n,f) sensor with cadmium cover:

room return (0.

source capsule (

support structures (

Ht Ie

238U(n,f) threshold sensor:
room return (0.0
source capsule (0.0
support structures (0.3

oe
ooo
wae
Loe
earns

Tall uncertainties are lo.

*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 252Cf spontaneous fission neutrons is similar to
that of 235Y and 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 252Cf spectrum is specified here together with the 235U
spectrum. A concise description of fission neutron spectra may be given in
terms of a broad energy range,

lower bound median upper bound
E, (p=0.99) E,(p=0.5) E,(p=0.01)
Pp P Pp
252¢f 0.09 MeV 1.68 MeV 7.8 MeV
235y 0.08 1.57 7.2


NBS MEASUREMENT SERVICES:
ACTIVATION FOIL IRRADIATION WITH
CALIFORNIUM 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


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
o = fission-spectrum-averaged reaction cross section (om?)
® = free-field fission neutron fluence (n/em*). (Note: the symbol

® is recommended by ICRU and ASTM E170.)
o°@ = 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):

A
—— =a + <> (3)
CaTCIN
A = saturated specific activity
CATCIN
[ATC] = decay correction factor for average neutron fluence rate. For
an uninterrupted irradiation of length,T at constant
fluence rate, [ATC] is equal to [1~-e d.
T = length of irradiation given in the test report.
<o> = the neutron fluence rate (n/em?s)
o<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 = YeA+C+Ne(o+o), (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

B= CAN/(AIHA)]+¥eA+CeNe (oe), (5)
where } is the decay constant of the parent activity and A' that of the
daughter.


(1)

[2]

[3]

[4]

[5]
[6]

[7]

[8]

[9]

[10]

{11}

[12]

[13]

(14)

REFERENCES

Grund], J., and Eisenhauer, C., "Compendium of Benchmark Neutron Fields for
Reactor Dosimetry," NBSIR 85-3151 (January, 1986).

Grund], J. A., Spiegel, V., Eisenhauer, C. M., Heaton II, H. T.,

Gilliam, D. M. NBS; and Bigelow, J., ORNL; "A Californium-252 Fission
Spectrum Irradiation Facility for Neutron Reaction Rate Measurements," Nucl.
Tech. 32, 315 (March, 1977).

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).

DeJuren, J. A., Padgett, D. W., and Curtiss, L. F., J. Res. NBS AO, 63
(1955).

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

Noyce, R. H., Mosburg, E. R., Ur., Garfinkel, J. B., and Caswell, R. S.,
“Absolute Calibration of the National Bureau of Standards Photoneutron
Source-III, Absorption in a Heavy Water Solution of Manganese Sulfate,"
Reactor Sci. Technol. (J. Nucl. Energy, Part A/B) 17, 313 (1963).

Spiegel, V., private communication,

McGarry, E. D. and Boswell, E. W., "Neutron Source Strength by the Manganese
Sulfate Bath Method," NBSIR, to be published.

Li Linpei, "Reflection of 252Cf Fission Neutrons from a Concrete Floor,"
Radiation Protection Dosimetry, 5, 237 (1984).

Grund], 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).

Grundi, J. A. and Eisenhauer, C. M., “Fission Spectrum 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
Commerce, Washington, D.C. (March, 1975).

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).

Madland, D. G., LaBauve, R. J., and Nix, J. R., "Differential and Integral
Comparisons of Three Representations of the Prompt Neutron Spectrum for the
Spontaneous Fission of 252Cf," Los Alamos National Lab. Report
LA-UR-84-3557 (1984).

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).

17


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 System-for 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.

Atomic and Plasma Radiation Division

The Center consists of five Divisions and one Group.

Carries out basic theoretical and experimental research into the
spectroscopic and radiative properties of atoms and highly ionized
species; develops well-defined atomic radiation sources as radiometric
or wavelength standards; develops new measurement techniques and
methods for spectra) analysis and plasma properties; and collects,
compiles, and critically evaluates spectroscopic data. The Division
consists of the following Groups:

Radiation Physics Division

‘@ Atomic Spectroscopy
@ Atomic Radiation Data
@ Plasma Radiation

Provides the central national basis for the measurement of far ultra-
violet, soft x-ray, and electron radiation; develops and disseminates
radiation standards, measurement services, and data for for these radia-
tions; conducts theoretical and experimental research with electron,
laser, ultraviolet, and soft x-ray radiation for measurement applica-
tions; determines fundamental mechanisms of electron and photon inter-
actions with matter; and develops advanced electron- and photon-based
measurement techiques. The Division consists of the following Groups:

Radiometric Physics Division

Far UV Physics
Electron Physics
@ Photon Physics

Provides national measurement standards and support services for ultra-

violet, visible, and infrared radiation; provides standards dissemination

and measurement quality assurance services: conducts research in optical
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

e@ Spectral Radiometry
@ Spectrophotometry
© Radiometric Measurement Services

Develops, operates, and improves major NBS radiation facilities
including the electron Linac and race track microtron; develops,
designs, and builds electronic and mechanical instrumentation for
radiation programs and facilities; provides national leadership
in the standardization of nuclear instrumentation; and develops
new competence in radiation sources and instrumentation. The
Division consists of the following Groups:

lonizing Radiation Division

Accelerator Research
Linac Operations
Electronic Instrumentation
Mechanical Instrumentation

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
jonizing 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:

Nuclear Physics Group

Office of Radiat ion Measurement

Radiation Theory

Radiation Chemistry and Chemical
Dosimetry

Neutron Measurements and Research

Neutron Dosimetry

Radioactivity

X-Ray Physics

Dosimetry

Engages tn 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.


APPENDIX C

TEST REPORT FOR NEUTRON FLUENCE STANDARDS

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):
a= n/om?

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

Additional parameters of irradiation:
Decay correction factor: C = using A =
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 (10)
(1) Neutron source strength of ?5?cfr 1.18
(2) Distance measurement for compensated geometry
(3) Neutron scattering including sensors

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:


READ(5,3)1ID,M,IYR3
CALL IDAYS(ID,M,1IYR3,IDAY3)
CALL LPDAYS(1YR3,1YR2,1TL)
N=( IYR3-IYR2)*365+IDAY3-IDAY2+ITL+1
GO TO 80
70 N=l
WRITE(7,23)
READ(5,3)1ID,M,IYR2
IF(ID.EQ.0)GO TO 200
CALL IDAYS(ID,M,1YR2,IDAY2)
80 CALL LPDAYS(IYR2,1YRO,ITL)
IT2=( IYR2-IYRO)*365+IDAY2-IDAYO+ITL
cx DEBUG OPTIONS FOLLOW
c WRITE(7,500) IDAYO, IYRO, IDAY1,1YR1,IDAY2,1YR2,ITL,IT2
C 500 FORMAT(7X,‘IDAYO,1IYRO,IDAY1,IYR1,IDAY2,1YR2,ITIL,IT2' ,/7X,816)
T=FLOAT(IT2)
DO 100 I=1,N
A50=A500A4EXP(-0.693147AT/ (CFOHFA365. 25) )
A52=A520xEXP(-0.693147AT/ (CF2HF*365.25) )
A54=A54 04EXP(-0.693147AT/CF4HF)
ATT=A50+A52+A54
AT1=ATT/AAT
FO1=A50/AIT
F41=A54/ATT
CN1=CNOAAT1
CALL MODAY(IYR2,IDAY2,A,ID)
IF ( (REP.EQ.‘Y’).OR.(REP.EQ.‘y’) ) THEN
WRITE(7,2)A,1ID,1YR2,F01,F41,AT1,CN1
END IF
WRITE(2,2)A,1ID,1YR2,F01,F41,AT1,CN1
IDAY2=IDAY2+1
IF(IDAY2.LT.366)GO TO 100
IF (IDAY2.GT.366)GO TO 95
IMOD=MOD( IYR2,4)
IF(IMOD.NE.0)GO TO 95
GO TO 100
95 IDAY2=1
TYR2=IYR2+1
100 T=T+1.
GO TO 60
200 CLOSE(2)
WRITE(7,21)
STOP
END

SUBROUTINE IDAYS(ID,M,IYR,IDAY)
THIS SUBROUTINE CALCULATES THE JULIAN CALENDAR DA¥ OF THE YEAR

qaqa ana

1 FORMAT(7X‘YOU TYPED MONTH AND DAY. TYPE DAY AND MONTH '/)
3 FORMAT(1I5)
IF(1IYR.LT.1900) IYR=IYR+1900
IF(M.GT.12)WRITE(7,1)
IF(M.GT.12)READ(5,3)ID,M
IY=MOD(IYR,4)
Cc FOR MONTHS AFTER FEBRUARY
X=2.0


Fig. 2. Typical arrangement of six activation sensors around the ?°*Cf source.


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 100.

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+[¥r2+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 over-
all 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 (1c). For our sample holders, which are not perfectly rigid, the
reproducibility of measurements is + .04 mm (1c). For the previously
discussed case of a sample 50 mm from the source (100 mm sample separation),
the uncertainty from positioning uncertainty is + .08% (lo).

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-distance measure-
ments 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 cylin-
drical symmetry, More extensive Monte-Carlo calculations have been carried
out for special sensors, specifically the NBS double fission chamber.