TABLE 3. ASSIGNMENT OF UNCERTAINTIES (+ 1 o)

e Fluence Transfer Measurements from the 252Cf Standard Fission
Neutron Field to the 235 Cavity Fission Source:

Percent

(1) Absolute neutron source strength of the 252Cf source 0.9 S#R
(2) Distance measurement 0.6 R
(3) 252cf to 235U spectrum-averaged cross section ratio

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

counting statistics, and electronic system stability

over period of measurement) 0.4 R

Subtotal (summed in quadrature) = 1.5
e Preparation of Neutron Fluence Standards at the

235 Cavity Fission Source:
(1) Gradient corrections and positioning 0.4 S+R
(2) Corrections for neutron scattering for 115In(n,n') 1.2 R
(3) Reproducibility of fluence monitoring including random

summing corrections and uncertainties in activation

decay corrections 0.4 R

Subtotal (summed in quadrature) = 1.3

TOTAL UNCERTAINTY = 2.0

Se,

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


aaqaa

aaa

a0

TO FILENAME.RES, IT WILL ADJUST THE PEAK PARAMETERS WITH RESPECT
TO THE PEAK FOR CHANGES IN GAIN OF THE PEAK. IT CALLS SUBROUTINE

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

DIMENSION EGY(40),IPK(4O),IWL(40), IWH(40) ,PK(4O),
&IC(4O),ID(40) ,NL(40) ,NR(40)
DIMENSION BLOW(40),BHIGH(40),BKG(40) , TOTAL (40) ,ER(40)
CHARACTER IFILI(20),IFILO( 20), IREP
LOGICAL*1 GAIN,PRMS, PULSE
DOUBLE PRECISION XME(4)
VIRTUAL Y(8192)
COMMON/ ADCAM/IADCAM
COMMON/IPUT/XME
COMMON/LAMBD/IYRO, IDAYO , IHRO,MINITO,SECO, ALAM, CASRAT,
&IYR, IDAY, IHR, MINIT, SEC, PRMS
COMMON /ENEG/ EGY
COMMON /SPEC/ TIME,CLTM, PLSRTE
COMMON /ENFIT/ A,Z,C
COMMON /PKLIM/ NUM, IWL,IWH,IC,ID,NL,NR,PK,IBW,SIG
COMMON /FILNMS/ IFILI,IFILO
DATA GAIN/.FALSE./
IF(IADCAM.GT.0)GO TO 200
OPEN ( UNIT=2 , FILE='KEVO.DOC' ,STATUS=" OLD")
GO TO 201
200 OPEN(UNIT=2,FILE='"KEV1. DOC! ,STATUS=' OLD")
201 READ(2,1)A,Z,C
FORMAT (3E17.10)
2 FORMAT(A1)
CLOSE(2)
TYPE 22
22 FORMAT('$DO YOU WANT TO CORRECT FOR GAIN SHIFT ? ')
ACCEPT 2,IREP
IF(IREP.EQ.'Y') GAIN = .TRUE.

OUTPUT RESULTS

TYPE 3

WRITE(4,3)
3 FORMAT (1 #®® 8H HR RH KK HHH KKK KKK KERR ERE ERE

BEI EI ET)

WRITE(4,4)IFILI
TYPE 4,IFILI

4 FORMAT(/* DATA FILENAME: '21A1)
WRITE(4,5)TIME
TYPE 5,TIME

5 FORMAT(/* LIVE TIME (SEC): ',F8.0,/)

21 CONTINUE

7 FORMAT(' PEAK CHANNELS ENERGY BKGND AREA
& CTS/SEC'/)

9 FORMAT(/)

DETERMINE AREAS

% ERR


°

Qaaqa0

eaa

10

20

30
4O

45

SUBROUTINE CAL(Y)

SUBROUTINE CAL USES SUBROUTINE ELOC TO FIND PEAKS
IT IS ONLY CALLED BY THE MAIN PROGRAM

DIMENSTON X(40),F(40)

VIRTUAL Y(8192)

COMMON /PEAK/X,E,N

COMMON/ ADCAM/IADCAM

DATA SX,SX2,SX3,SX4,SY,SXY,SX2Y/7*0.0/
CALL ELOC(Y)

TYPE 2
FORMAT(' PEAK CHANNEL ENERGY '/)
DO 11 I=1,N

TYPE 12,1,X(I),E(1)
CALCULATE PEAK ENERGY : E = A + BX + CX¥#2

FORMAT(I4,5X,F9.2,2X,F9.2)
IF(N.EQ.1) GO TO 20

DO 10 I=1,N

SX=SX+X(I) ! SUM X'S
X2=X(1)*X(T)

SX2=SX2+K2 ! SUM X**2
SX3=SX3+X2*X(I) 1 X¥*3
SXU4=SX4+X2*XQ 1 SUM X¥X4
SY=SY+E(I) ! SUM Y'S

SXY=SXY+X(I)*E(I)! SUM X*Y
SX2Y=SX2Y+X2*E(1I)! SUM Y*X**2

CONTINUE

T=N

CKSX3=T*SX3-SX*SX2 ! THIS IS A CLOSED SOLUTION TO EQ (8-4)
CKSX4=T*SX4-SX2*SxX2 ! OF BEVINGTONS'S DATA REDUCTION AND ERROR
CKSX2=T*SX2-SX*SX ! ANALYSIS FOR THE PHYSICAL SCIENCES

CKSXY=T*SXY-SX*SY
CKSX2Y=T*SX2Y-SX2*SY

C1 =CKSX2*CKSX2 ¥-CKSX3*CKSXY
C2=CKSX2*CKSX4-CKSX3*CKSX3
c=C1/C2

B1=CKSXY-C*CKSX3

B=B1/CKSX2

A1=SY-C*SX2~-B¥*SX

A=A1/T

GO TO 30

A=0.0

C=0.0

B=E(1)/X(1)

TYPE 40

FORMAT(/' DATA FILE KEV.DOC = '/)
TYPE 45,A,B,C

FORMAT(' A, B, & C = ',1P3E14.7)

IF(IADCAM.GT.0)GO TO 200


Appendix E

PROGRAM SCATFO TO CALCULATE SCATTERING CORRECTION FOR ACTIVATION
FOILS IN CAVITY FISSION SOURCE. DATA MAY BE ENTERED FROM A DISK FILE
OR FROM THE KEYBOARD
INPUT DATA:
ANAME(J)= FOIL ID
N = THE NUMBER OF FOILS
IT(J) = FOIL TYPE
POS(J)= POSITION OF FOIL IN MILS
TH(J) = THICKNESS OF FOIL IN MILS
CHARACTER*4 ANAME(50)
DIMENSION GRAD(50)
DIMENSION IT(50),POS(50),TH(50)
OPEN (3,STATUS="NEW')
TYPE 97
97 FORMAT(1X,'ENTER FILE NUMBER FOR DATA; 5 FOR KEYBOARD')
READ(5 ,100)NF
100 FORMAT(I10)
IF(NF.NE.5)GO TO 101
TYPE 99
99 FORMAT (1X,'ENTER NUMBER OF FOILS, C.R.')
GO TO 104
101 OPEN(NF ,STATUS=" OLD! )
104 READ(NF,100) N
IF(NF.NE.5)GO TO 103
TYPE 98
98 FORMAT(1X,'ENTER FOIL ID IN AY FORMAT FOLLOWED BY TYPE,'
&,/' POSITION IN MILS, THICKNESS IN MILS, FOLLOWED BY'/
&' C.R. AND THEN NEXT FOIL. USE 1 FOR IN, 2 FOR NI AND 3 FOR AL")
103 READ(NF,110)(ANAME(J),1IT(J),POS(J),TH(J),J=1,N)
WRITE(7,111) (ANAME(J),IT(J),POS(J),TH(J),J=1,N)
110 FORMAT(A4,110,2F10.1)
111 FORMAT(1X,A4,110,2F10.1)
WRITE(3,96)
96 FORMAT(2X,'ID',6X,'GRAD.',8X, 'TYPE',2X, 'POSITION' ,2X, ' THICKNESS
& FOIL SCAT MUSC')
DO 1000 I=1,N
GRAD(I)=1+21,2E-6*POS(I)*POS(I)
SCAT=0
IF (IT(I)-2)400 ,500 ,600
400 DO 450 J=1,N
IF(I.EQ.J)GO TO 450
X-ABS(POS(J)-POS(I))
-001+(,0054*EXP(~.0728*X) )
Y=Y*TH(J)/10.
120 FORMAT(1X,F10.3)
SCAT=SCAT+Y
450 CONTINUE
ALSCAT=SCAT+ .0046+(.00255*TH(I)/5.)
Go TO 800
500 DO 550 J=1,N
IF(I.EQ.J)GO TO 550
X=ABS(POS(J)-POS(I))

E-1


TABLE 1. EVALUATED FISSION NEUTRON SPECTRA FOR 252Cf AND 235y
N 45-GROUP FORMAT

252¢f 235y
Lower
Energy Group Cumul ative() 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 0.9615
0.25 0.0152 0.9528 0.0166 0.9461
0.3 0.0323 0.9377 0.0350 0.9295
0.4 0.0337 0.9054 0.0363 0.8945
0.5 0.0343 0.8717 0.0367 0.8582
0.6 0.0343 0.8374 0.0365 0.8215
0.7 0.0338 0.8031 0.0358 0.7851
0.8 0.0664 0.7693 0.0697 0.7493
1.0 0.0641 0.7029 0.0668 0.6796
1.2 0.0608 0.6388 0.0629 0.6128
1.4 0.0290 0.5780 0.0298 0.5499
1.5 0.0279 0.5490 0.0286 0.5201
1.6 0.0527 0.5211 0.0536 0.4915
1.8 0.0484 0.4684 0.0487 0.4379
2.0 0.0442 0.4200 0.0441 0.3892
2.2 0.0206 0.3758 0.0204 0.3451
2.3 0.0196 0.3552 0.0193 0.3247
2.4 0.0364 0.3356 0.0356 0.3054
2.6 0.0328 0.2992 0.0319 0.2698
2.8 0.0296 0.2664 0.0284 0.2379
3.0 0.0503 0.2368 0.0477 0.2095
3.4 0.0310 0.1865 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
W 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 3gne shape function ER. exp(-aE) ;
a = 0.70 for 252¢f; = 0.76 for 235U,

(8) spectrum fraction above lower energy bound.


Recommended Procedure for Establishing a Calibration Factor (x,5)

Use Equation (1) and either Equation (2) or Equation (3) to derive a
free-field neutron fluence or average fluence rate, respectively, from the
measurement of a particular activity of the neutron sensor. Define a
calibration factor for this activity as the ratio of the NBS certified
fission neutron fluence or fluence rate to the derived neutron fluence or
fluence rate. Then, to obtain an NBS calibrated result for subsequent
measurements with this activity, either multiply observed reaction
probability by the calibration factor or divide the detector cross section
by the calibration factor. This calibration procedure will be most
effective if the derived fluence or fluence rate for the NBS Standard is
obtained with the same reaction parameters, and with the same fission
spectrum shape and detector cross section set as are used in the transport
calculations which provide spectrum-averaged detector cross-sections for

subsequent neutron field measurements.

Revised: November 3, 1986

G-4


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

Irradiation (EOI) for Cavity Fission Neutron Source Irradiations

The neutron fluence or fluence rate standard has been exposed to a fission
neutron fluence that is perturbed by a small cavity return flux and by the
effects of neutron scattering in fission source disks, encapsulations, and
the fluence standards themselves. This test report certifies a free-field
fission neutron fluence or fluence rate, i.e. corrected for neutron
scattering. This correction, based on calculation, is reported as the
fractional departure from the free-field activation attributable to neutron
scattering including cavity return. On. this basis the total activity
measured at EOI is

ALT AC + uy) qa)

where: Aa = total activity measured at EOI
A = net free-field activity at EOI
Uso = fractional departure from free-field activity due to
neutron scattering. (A minus sign means that the observed
activity is lower than it would be in free-field conditions

because of a net outscatter of neutrons.)

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

activity of the neutron fluence standard in disintegrations per second is

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

A-c-e (2)
ACN
where: A = net free-field activity at EOI (dps)
dX = decay constant of reaction product
C = decay correction factor given in test report. For an
uninterrupted irradiation of length T at a constant fluence
rate, C is equal to [(1-exp(-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 (om2 )

G-2


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

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

[16] Madland, D. G., and Nix, J. R., “New Calculation of Prompt Fission
Neutron Spectra and Average Prompt Neutron Multiplicities," Nucl. Sci.
Eng. 81, 213-271 (1982).

[17] 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).

[18] Grund], J. A. and Eisenhauer, C. M., "Fission Spectrum Neutrons for Cross
Section Validation and Neutron Flux Transfer," Proceedings of a
Conference on Nuclear Cross Sections and Technology, NBS Special Publi-
cation 425, U.S. Dept. of Commerce, Washington, D.C. (March, 1975).

[19] Wiedling, T., Proc. Symp. on Neutron Standards and Flux Normalization,
ANL (1970); see also Smith, A. B., Proc. Consultants Meeting on Prompt
Fission Neutron Spectra, Vienna (1971); Koster, A., Ibid., Vienna
(1971).

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

[21] Magurno, B. A., Kinsey, R. R., and Scheffel, F. M., Guidebook for the
ENDF/B-V Nuclear Data Files, BNL-NCS-31451 (1982).

[22] Lamaze, G. P., Gilliam, D.M., and McGarry, E.D., "Neutron Fluence and
Cross Section Measurents for Fast Neutron Dosimetry," Proc. Conf.
Nuclear Methods and Environmental and Energy Research, Puerto Rico,
(April, 1984).


IDAYS = 0
SECS = 0.0
IF(.NOT. PRMS) GO TO 72

10 IF(IYRO.LE.IYR)GO TO 70
WRITE(5 ,12)1YRO,IYR

12 FORMAT(' IYRO,',I5,', CANNOT BE LARGER THEN IYR,',15)
TYPE 18

18 FORMAT(' INPUT NEW IYRO ')
ACCEPT 19, 1YRO

19 FORMAT(1I6)
GO TO 10

70 IF(IYRO.EQ.IYR)GO TO 71
IF(MOD(IYRO,4).EQ.0)IDAYS =IDAYS+1
IYRO=IYRO+1
IDAYS=IDAYS+365
Go TO 70

71 IDAYO=IDAYO-IDAYS
FDAYS=IDAY-IDAYO
FIHRS=FDAYS*24 . + (IHR-IHRO)
FMINIT=FIHRS*60. + (MINIT-MINITO)
SECS=FMINIT*60 . + (SEC-SECO)

72 TYPE 88,SECS
WRITE (4 ,88)SECS
WRITE(4,7)
TYPE 7

88 FORMAT(' SECONDS AFTER END OF IRRADIATION = ',E14.6,/)
PULSE = .FALSE.
IF(PLSRTE.LT.2) GO TO 885
PULSE = .TRUE. ! PULSER PEEK IN SPECTRUM
NNUM=NUM+1
DO 870 I=NNUM,2,-1
EGY (I)=EGY(I-1)
IPK(1)=LPK(1-1)
IWL(I)=IWL(1I-1)
IWH(I)=IWH(I-1)
PK(I)-PK(I-1)
Ic(1)=IC(I-1)
ID(I)=ID(I-1)
NL(I)=NL(I-1)

870 NR(I)=NR(I-1)
EGY(1)=EGY(NNUM)
IPK(1)=IPK(NNUM)
IWL(1)=1WL(NNUM)
IWH(1)=IWH(NNUM)
PK(1)=PK(NNUM)
I¢(1)=I¢(NNUM)
ID(1)=ID(NNUM)
NL(1)=NL(NNUM)
NR(1)=NR(NNUM)

GO TO 886
885 PLSRTE=1 ! USED IN ST 86
PLSRAT=1 ! USED IN ST 86
886 DU 40 I=1,NUM
IDIFF=0

D-12


TABLE OF CONTENTS

1, DESCRIPTION OF SERVICE. 2 ee ee eee eee ee ne
Configuration and Characteristics .......

1.1.

2. NEUTRON FIELD CHARACTERIZATION. 2. ee eee eevee
Neutron Fluence . 2... eee eee eee eee
Neutron Spectrum... ee ee ee ee ee eee
2.2.1. Calculation . 2... 2... 2 2. ee ee eee eee
2.2.2. Measurement . 2. ee eee eee eee
2.2.3. Evaluated Spectrum... .. 2.2.2. eee

2.1.
2.2.

(multigroup format, spectrum errors)

3. IRRADIATION PROCEDURES 2... eee eee eee eee

4. DETAILED OPERATING PROCEDURES .......
Assembly and Irradiation of Sensors .
Fluence and Fluence Rate Calculations.
Summary of Uncertainties .......

4.1.
4.2.
4.3.

5. BENCHMARK REFERENCING «2 2 ee ee eee eer ene

6. SAFETY © we eee ee eee ee et eee ee tn es

REFERENCES 6 6 we wee ee ee eee ee we eee eee

APPENDIX
APPENDIX
APPENDIX

APPENDIX
APPENDIX
APPENDIX
APPENDIX

A.

B.

Typical detector assembly worksheet .....

Typical irradiation log and fluence calculation worksheet

Operating instructions for gamma spectrometer
height analyzer... ee eee ee ee eee

Spectrometer data analysis program - STP2D8 .
Scattering correction program - SCATFO ...
Saturation factor program - SAT260 .....

Typical Test Report... 2... ee ee eee

pulse

ne

DADO

10
10
il
12
14
15
16
A-1

B-1

C-1
D-1


ACCEPT 2,REPL
IF(REPL.EQ."Y') GO TO 20
IF(REPL.EQ.'D') GO TO 21
IF(REPL.EQ.'H') GO TO 22
IF(REPL.EQ.'M') GO TO 23
IF(REPL.EQ.'S') GO TO 26
TA=TA¥*365.25

TA-TA*2N,

TA=TA*60.

TA=TA*60.
IF(TB.LT.1.0E-5) GO TO 50
TYPE 28

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

30
31
32
33
50
51

52
54

53
55

63
59
61

65

100

10

102

ACCEPT 2,REPL
IF(REPL.EQ.'Y') GO TO 30
IF(REPL.EQ.'D') GO TO 31
IF(REPL.EQ.'H') GO TO 32
IF(REPL.EQ.'M') GO TO 33
IF(REPL.EQ.'S') GO TO 50
TB=TB*365.25

TB=TB¥24,

TB=TB*60.

TB=TB*60.

TYPE 51

FORMAT(' HOW MANY PEAKS DO YOU WISH TO USE (12) ")
ACCEPT 53,NUM

TYPE 52

FORMAT(/'$PEAK SIGNIFICANCE : ')

ACCEPT 54,SIG ! USED AFTER ST 70 IN SUB CHPEAK
FORMAT (F6.0)

FORMAT (16)

FORMAT(5X,' ENTER THE LEFT BKG. CHANNEL, PEAK CHANNEL AND',/,
&! THE NUMBER OF BKG. CHANNELS TO BE AVERAGED ON THE LEFT. ENTER',/
&' UPPER PEAK CHANNEL ,THE BKG. CHANNEL, AND NUMBER OF CHANNELS ',/
&' TO BE AVERAGED ON THE RIGHT. FOLLOW THIS WITH THE PEAK POSI-',/
&' TION. ALL ENTRIES IN 61, 1F FORMAT ON ONE LINE WITH A COMMA',/
&' BETWEEN EACH.')

TYPE 55

DO 63 I=1,NUM

TYPE 59,1

ACCEPT 61, IWL(I),1C(I),NL(I),ID(1),IWH(I),NR(I),PK(I)

FORMAT(7X," PEAK NO. ',I4)

FORMAT (6110,F14.0)
WRITE(2,104)NUM, IYRO, IDAYO, IHRO ,MINITO,SECO,SIG,TA,TB

DO 65 I=1,NUM

WRITE (2,105 )IWL(I) ,IC(1) ,NL(),ID(I),IWH(I),NR(I),PK(I)
CLOSE(2)

GO TO 120

OPEN( UNIT=2, FILE='"LIMITS.DOC' ,STATUS="OLD! )

TYPE 101

FORMAT('$WHAT IS THE ISOTOPE KEYWORD ? ')
ACCEPT 102, KEY

FORMAT(A8) ! DOUBLE PRECISION = 8 CHARACTERS
WRITE(5 122) KEY


5. 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 detector to a known neutron fluence, Xo5° ina
235y fission spectrum establishes a calibration factor,

¢ - [observed detector response] (1)
[xy5 neutron fluence]

which may be used in conjunction with the calculated fission-spectrum-averaged
cross section, Oy, for the detector reaction to obtain a consistent benchmark

calibration, The cross section, Os 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 og and o; (commonly from the ENDF/B-V dosimetry file).

[If the absolute detection efficiency for the dosimetry measurement
method has been established (e.g., a good resolution fission chamber with
known deposit mass, or a calibrated gamma-ray counting system and known
reaction parameters, branching ratios, isotopic abundances, fission yields
etc.), then the “observed detector response" in Eq. (1) becomes the reaction
probability, and G is an observed cross section for the fission spectrum.]

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

By irradiating samples nearly identical to those irradiated in the 235y
field, the fluence of the study field, 6, tudy> can be obtained as a simple
ratio of detector responses, D, and spectrum’ averaged cross sections, 5, so
that:

study _ “study “study (2)
D 3 ®
x °% xX

14


Appendix B

IRRADIATION LOG AND WORKSHEET FOR CALCULATING NEUTRON FLUENCE

Irradiation Identification:

Materials Irradiated:

id.
Ni-3
Ni-4
Ni-5

Ni-6

Source Assembly

Top Source Disk:

Bottom Source Disk:

8601

Au, Ni, Se

Source Separation Distance: 443

Irradiation Log:

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

EOI: 21:37:00 EST Jan 20, 1986

Duration 2 days. 10 hr, 36 min

Fission chamber monitor:

Not used

R/hr @ 10cm

(DAY 18 )

(DAY 20)

= 2109608

Nickel and Indium Fluence Monitor Response at Midplane

+15

+O

4(z)/0(0)

1.0178
1.0010

1.0048

1.0339

5491.1
5395.1
5368.0

5436.0

v(c/s,g) ‘> activation by
@ BOL

D(e/s,g)
@ EOL
Scattered Flux (corrected)
-0.007 5433
-0.007 5428
-0.007 5380
-0.007 5295

5385433 (0.6%)


(1)

[2]

[3]

4]

(5]

[6]

[7]

[8]

[9]

[10]

[11]

[12]

[13]

REFERENCES

Deduren, J. A., Padgett, D. W., and Curtiss, L. F., J. Res. NBS 40, 63
(1955).

"NBS Calibration Services, Users Guide 1986-1988." NBS SP250, U.S.
Government Printing Office, Washington, D.C. (1986).

McGarry, E. )., Lamaze, G. P., Eisenhauer, C. M., Gilliam, D. M., and
Schima, F. J., "NBS ISNF and Cavity Fission U-235 Standard Neutron
Fields," Proc. Fourth ASTM-EURATOM Symposium on Reactor Dosimetry,
Washington, D.C., NUREG/CP-0029, Vol. 2, 597 (March, 1982).

McGarry, E. D., Eisenhauer, C. M., Gilliam, D. M., Grund], J., and
Lamaze, G. P., "The U.S. U-235 Fission Spectrum Standard Neutron Field
Revisited," Proceedings Fifth ASTM-EURATOM Symposium on Reactor
Dosimetry, Geesthacht, Germany (September, 1984).

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).

Eisenhauer, C. M. and Grund], J. A., “Neutron Transport Calculations for
the Intermediate-Energy Standard Neutron Field (ISNF) at the National
Bureau of Standards," Proceedings International Symposium on Neutron
Standards and Applications, NBS Special Publication 493, U. S. Dept. of
Commerce, Washington, D.C. (March, 1977).

Fabry, A. and Jenkins, J. D., "Wall Return Neutron Fluxes for High and
Intermediate-Energy Cavity Neutron Sources," Trans. Am. Nucl. Soc.
15, 2 (1972).

Fabry, A., DeLeeuw, G., and DeLeeuw, S., CEN/SCK, Mol, Belgium, "The
Secondary Intermediate-Energy Standard Neutron Field at the Mol-r5
Facility," Nucl. Tech. 25, 349 (February, 1975).

Fabry, A., Minsart, G., Cops, F., and DeLeeuw, S., “The Mol Cavity
Fission Spectrum Standard Neutron Field and Its Applications," Proc.
Fourth ASTM-EURATOM Symposium on Reactor Dosimetry, Washington, D.C.,
NUREG/CP-0029, Vol. 2, 665 (March, 1982).

Grund], J., "Cavity Fission Spectra," Trans. Am. Nucl. Soc. 5, 381
(1962). ~

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.

16


c
Cc

122
103

104

106
105

110
112
120

FORMAT (1X, A8)

READ(2,102,END=110) TOKEN

IF(TOKEN.NE.KEY)GO TO 103

FORMAT (518,F8.0,F8.0,/,2E16.6)
READ(2,104)NUM, IYRO, IDAYO, THRO, MINITO,SECO,SIG,TA,TB
WRITE(5 ,104)NUM, IYRO, IDAYO, IHRO,MINITO,SECO,SIG,TA,TB
DO 106 I=1,NUM
READ(2,105)IWL(1),1C(1),NL(1),1D(1),IWH(1),NR(1),PK(1)
WRITE(5,105)IWL(I),1C(I),NL(1),ID(I),IWH(I),NR(I),PK(I)
FORMAT (618 ,F8.2)

CLOSE(2)

GO TO 120

TYPE 112

FORMAT(' KEY WAS NOT FOUND IN LIMITS.DOC FILE ')
ALAM=ALOG(2.)/TA

CASRAT=(TA-TB)/TA

RETURN ! TO MAIN PROGRAM

END ! OF SUBROUTINE PARAMS

333 G GOGO BEGET a ae

c
c

(OHH B GOUGH OOOOH O UFO G HOHE FGI GEE ee

c

aa

aa

SUBROUTINE GETFIL(LUN,AREAI ,AREAO)

THIS SUBROUTINE IS CALLED FROM MAIN IN ORDER TO ASSIGN
THE INPUT AND OUTPUT FILES TO LOGICAL UNITS

CHARACTER*20 AREAI, AREAO
TYPE 1 .
FORMAT('$ENTER INPUT FILENAME ')
READ(LUN,9) IQI, AREAI(:IQT)
IQ=IQI

K=0

DO 10 I = 1,1Q

IF (AREAI(I:1) .EQ. '.") K
IF (K.EQ.O)AREAI(IQ+1:1Q+4
IF(K .NE. 0) IQ=K~1
AREAO(:1Q) = AREAI(:1Q)
AREAO(IQ+1:1Q + 4) = '.RES?

FORMAT(Q, A20)

WRITE (LUN ,8)AREAI(:1Q+4) ,AREAO( :1Q+4)

FORMAT (2X, A20,3X,A20)

OPEN(UNIT=3, FILE=AREAI (:1Q+4),STATUS=' OLD' ,ACCESS="DIRECT',
1 RECORDSI ZE=32)

OPEN(UNIT=4, FILE=AREAO( :1Q+4) , STATUS="NEW' )

RETURN ! TO MAIN PROGRAM

END ! OF SUBROUTINE GETFIL

I! DECIMAL POINT
~SPC

iret ttt t tert tert i ttt ttt ttt ttt ttt t eter rt irtt iti eetrrerrrrerrr rere rers ee

c

(00H OHO G OGG GGE GGG GGG CR Re

D-6


4.1.3. Arrange time at the thermal column with other users, principally
the radiography group and the activation analysis group.

4.1.4, Source disk assembly. The detector capsule is now ready for
assembly with the source disks. Two pairs of source disks are presently in
use (CV-3 through CV-6). Each source disk (15.9-mm dia. x 0.15-mm thick) is
made up of two 0.08-mm thick enriched uranium metal foils wrapped in 0.04-mm
thick aluminum foil, The total weight of a wrapped source disk is about
0.60 g and the total thickness is 0.38 mm. During irradiation the fission
neutron source strength of the two disks is about 4 x 101! n/s.

When not in use the source disks are kept at the far end of the lead
shield cart inside of the lead cup (LC). (To get LC in and out remove lead
bricks and short cover plate at front of cart.) Always check the radiation
level of the source disks before handling. Use the lead cup as a carrier when
moving source disks to the assembly area. Wear glasses for beta shielding.

4.1.5. Set up fission chamber monitor and electronics. Run “timer"
program on HP85 (see Appendix C). Check time marking instrument with
telephone time (844-2525) or Naval Observatory Time (653-0258), Notify
reactor control room. Raise the boral curtain while noting the time. Record
the time in notebook, irradiation log and worksheet (see appendix B), and
thermal column log sheet.

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

4.1.7. At the end of the irradiation (EOI), again check time. Notify
reactor control room. Lower the boral curtain while noting the time. Record
the EOI time in notebook, worksheet, and thermal column log.

4.1.8. Remove sensors from cadmium detector capsule using gloves and
package individually in marked plastic containers. Store in shielded card
file in building 235, room B123. Aluminum spacers and detector capsule should
be stored in pig either in room B123 or in thermal column area. Cadmium box
should be returned to its plexiglass holder and also stored as appropriate to
its residual radiation level.

4.2. FLUENCE AND FLUENCE RATE CALCULATIONS

4.2.1. Monitor Counting. The gamma radiation from the nickel (or
indium) foils fs counted in the standard geometry with the intrinsic germanium
detector. The standard geometry is foil holder #3 mounted directly on the
detector with the foil located in the recessed hole. A pulser should be fed
into the preamplifier and the rate (usually 50 pps) adjusted to obtain suffi-
cient counts in the pulser peak for the chosen counting interval. Detailed
instructions for the operation of the counting system are in appendix D, The
stored spectra are analyzed with the code STP2D8, (a FORTRAN code). A listing
of this code is in appendix E. The code includes a time history correction
and gives a final result as a counting rate at EOI. All results should be
corrected for deadtime and pileup by finding the live-time correction which

ll


NBS CAVITY FISSION SOURCE

235y sources

DETECTOR STACK

CADMIUM ENCLOSURE.

NESE
> Q

LA
NS \edaoe soe i)

CAVITY FISSION SOURCE
AND CADMIUM-COVERED

FREQ’ DV SS

Fig. 1. 235y Cavity fission source. Upper view: retative positions
of neutron sensor foils, fission disks and cadmium enclosure. Lower
view: the assembly in the thermal column of the NBS reactor.

jy



c
D
21

FORMERLY END ! OF SUBROUTINE READFI
WRITE (4,5)
TYPE 200

200 FORMAT(' | ENTER ADCAM NUMBER(O OR 1)')

9

13

12

15

25

Cc THI
65
105
28

2T

26
30

CHEEK

Cc
Cc

aaaaaaa

ACCEPT 4, TADCAM
IF (IADCAM.NE.O .AND. IADCAM.NE.1)GO TO 21

TYPE 9
FORMAT(/' THE ENERGY CALIBRATION DATA SHOULD BE ON KEV.DOC'//
&'$DO YOU WANT A NEW ENERGY CALIBRATION ? : ')

ACCEPT 2, REP
IF(REP.EQ.'Y') CALL CAL(Y)
TYPE 13
FORMAT(' DO YOU WANT TO LOCATE PEAKS, USING THE ENERGY '
&//'$CALIBRATION FROM KEV.DOC ? ')
ACCEPT 2, REP
IF(REP.EQ.'Y') CALL LOCATE(Y)
TYPE 12
FORMAT('$DO YOU WANT TO ANALYSE THE SPECTRUM : ')
ACCEPT 2, REP
IF(REP.EQ."Y') CALL EPEAK(Y)
TYPE 15
FORMAT(' DO YOU WANT TO INPUT ANOTHER FILE USING THE SAME’,
&/'$ PARAMETERS AND INPUT FORMAT ? ')
CLOSE (3)
CLOSE (4)
ACCEPT 2, REP
IF(REP.EQ.'Y') GO TO 10
TYPE 25, IFILO
FORMAT(7X'THE RESULTS ARE ON FILE ',20A1)
IF (TEM) GO TO 27
OPEN(UNIT=2,FILE='LIMITS.PKS' ,STATUS="OLD')
IE PREVIOUS LINE MODIFIED FROM RT-11 VERSION WHICH USED CALL ASSIGN
DO 65 I=1,NUM
WRITE(2,105)IWL(I),IC(I),NL(1),ID(1),IWH(I),NR(I),PK(I)
FORMAT (618,F8.2)
CLOSE(2)
TYPE 28
FORMAT(' THE PEAK LIMITS HAVE BEEN WRITTEN ON LIMITS.PKS.')
GO TO 30
TYPE 26
FORMAT(7X'DO NOT FORGET FILE LIMITS.TEM ')
STOP

END ! OF MAIN PROGRAM
RK RK KH KK RR IR III RRR KK RE ERK

SUBROUTINE PARAMS(TEM)

THIS SUBROUTINE PROVIDES THE INPUT FROM THE KEYBOARD OR FROM FILE
LIMITS.DOC FOR READING THE VARIOUS PARAMETERS FOR THE PROGRAM.
THOSE SUPPLIED FROM THE KEYBOARD WILL BE WRITTEN TO FILE LIMITS.TEM
SO THAT THEY CAN BE CONCATENATED TO LIMITS.DOC LATER.

IF ONE IS NOT INTERESTED IN PARAMETERS THE PROGRAM WILL

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


Appendix C

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

2.

3.
3.1
4,
5.
6.

14,
14,
1A,
14
qh.
wae

15.

16.

LOAD ADCAM MENU BY TYPING @MENU OR RUN DIRECTLY BY TYPING @START AND SKIP
TO STEP 3.1
THE "A" MENU WILL BE DISPLAYED; A CARRIAGE RETURN WILL CHANGE TO THE
"B" MENU; ANOTHER ONE CHANGES YOU BACK
FROM THE "B" MENU, TYPE "3" OR "START"

ANSWER QUERIES ABOUT MCA#
WHEN IT ASKS FOR PARAMETERS, TYPE LT
THEN TYPE IN THE PRESET TIME IN SECONDS
WHEN IT ASKS IF YOU WANT TO TRANSFER PARAMETERS FROM ANOTHER FILE,
SELECT NO (DEFAULT IS NO)
WHEN IT ASKS IF YOU WANT TO ALTER SAMPLE DESCRIPTION, TYPE Y
ENTER TWO LINE DESCRIPTION (DETECTOR DESCRIPTION 1S ALREADY ENTERED
IN ANOTHER PLACE)
MAKE SURE SAMPLE IS IN PLACE!
WHEN IT ASKS IF YOU WANT TO ALTER ANALYSIS PARAMETERS, TYPE N OR CR
(DEFAULT)

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

TAKING WILL END AT PRESET

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

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

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

TOOLKIT. FOR ADCAM 0 ANSWER ALL QUERIES WITH CARRIAGE RETURNS.
TO STORE DATA, ENTER EMULATE MODE (STEP 13)
1 HIT MENU KEY

SELECT SETUP MCA PARAMETERS

SELECT ADCAM # AND SEGMENT # (1)

WHEN DONE, HIT EXIT, SELECT SAVE PRIMARY SPECTRUM, GIVE NAME, DO
WHEN DONE, HIT EXIT TO RETURN TO STEP ONE FOR NEXT SPECTRUM
PLEASE UPDATE GEMLOG.DOC WITH NAMES OF STORED SPECTRA. PUT NAME OF
FLOPPY IN COMMENT SECTION IF NOT ON HARD DISC.

2
.3 SET PRIMARY SOURCE TO MCB
4
5


LIST OF FIGURES

Fig. 1 235U Cavity fission source .. 1.2.2.2 eee ee eee

Fig. 2 The NBS evaluation of the 235 fission neutron spectrum

vii


LIST OF TABLES

TABLE 1. Evaluated Fission Neutron Spectra for 252Cf and 235U in
45-Group Format... 1 ee ee ee ee te ee we ee ee ee 7

TABLE 2. Uncertainty Estimates for 252cf and 235U Evaluated Fission
Neutron Spectra 2... eee ee eee eee eee eee eee 9

TABLE 3. Assignment of Uncertainties . 2... ee eee ee ee ee ee 13

vi


would restore the pulser results to the known, true pulser rate. The deter-
mined counting rate must be corrected for losses of events due to attenuation
of the outgoing gamma rays. The correction factor, C. is equal to

(Leet) ut where y is the mass attenuation coefficient, for example, for the
810 keV line in nickel, » is 0.589 cm-!, The resulting correction for a
.0254 cm nickel foil is 0.9925. This corrected result is then divided by the
mass of the foil and entered as D in the worksheet,

4.2.2, Scattering corrections and gradient corrections, The FORTRAN
code "SCATFO" (see appendix F) is run on a Pro350 computer. Input data are
taken from the foil assembly worksheet and the program gives us,, the activa-

tion by scattered neutrons, The code also gives the field gradient,
¢(z)/o(0). These parameters should be entered on the fluence calculation
worksheet. A value of D at the midplane of the foil holder can now be
calculated for each foil and an average value obtained for the set of foils.

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

is run to obtain corrections for loss of activity during irradiation. If the
power level was approximately constant, a good correction can be obtained from

C= (1-e7At)/at where A is the decay constant and t is the length of the
irradiation, The activation decay rate factor, G, is equal to Cy’. The

neutron midplane fluence is now calculated using equation 4 of the worksheet.
Table 6 of the worksheet can now be completed which gives the fluence and the
scattering correction for each sensor. Test reports should also be prepared
at this point for each sensor. A sample test report is in appendix H.

4.3. SUMMARY OF UNCERTAINTIES

The uncertainties of the certified fluence are given in Table 3. The
first part of the table gives uncertainties from the fluence transfer
procedure from the 252Cf field to the 235y field. The second part of the
table gives uncertainties for preparation of the fluence standards at the
cavity fission source. The resultant total uncertainty is for the free field
fluence, Additional uncertainties due to scattering corrections in the
fluence standard must also be added. This last term varies depending on the
activation reaction and the placement of the sensors in the detector capsule.

The table indicates which uncertainties are mostly systematic and which
are mostly random. In some cases, the uncertainty terms consist of both
random and systematic components, A more important distinction in the
uncertainty terms is whether the uncertainty could lead to an overall change
in scale of the NBS neutron fluence values. Two components could have this
effect: 1) the measured value of NBS I could change (thereby changing all
source strength values, and 2) improved corrections. for neutron scattering for
Pu(n,f) in the Cf and 235 fission fields.

12


where: (A/aCTN)

a"

reaction rate probability (disintegrations per
second/nucleus.),

(0/T), the effective neutron fluence rate or flux density
(n/cm? s), and

2
"

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

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

The NBS cavity fission source operates at the center of a 30 cm diameter
spherical cavity in the NBS Research Reactor graphite thermal column, The
source-detector capsule consists of two coaxial source disks of 235U metal
(16 mm dia x 0.13 mm thick) placed outside a cadmium box that encloses passive
detectors for exposure. The cavity and source-detector arrangement are shown
in Fig. 1. The source-detector assembly is designed so that the cadmium box
and detectors may be removed separately from the fission disks. Fission
neutron fluence rates of ~ 4 x 1019 n/cm2 s are obtained at the midplane
between the source disks when they are at a separation distance of 11 mm. For
thin disk-shaped detectors with diameters up to 13 mm, axial fluence gradients
are mild: the fluence at 1 mm from the midplane is 4% greater than at the
midplane, Center-to-edge fluence ratios are more severe — see Section 2.1.

Fission neutron return for spherical cavities in graphite has been
studied extensively by means of neutron transport calculations and by experi-
ment in connection with the general development of standard neutron fields in
a spherical geometry [6]-[10]. For the 30-cm cavity at NBS, the response of
threshold fission detectors to fission neutrons returning from the cavity
walls is less than 0.3% of the uncollided fission neutron fluence from the
fission source disks; for detectors with thresholds above 1 MeV, the response
to the cavity return fluence is less than 0.1%. These estimates, based on
calculation, have been checked experimentally by means of fission chamber
traverses in the cavity fission sources at Center for Nuclear Studies in Mol,
Belgium (SCK/CEN) [9]. More important is fission neutron scattering in the
source-detector capsule, The scattered neutrons have a different energy
distribution then the original spectrum and therefore will generally have a
different interaction probability. Monte Carlo calculations are employed for
this problem and they provide corrections for threshold detectors which vary
with detection threshold and with the number and arrangement of detector disks
[3]. Corrections for activation rates are generally between 0.5% and 3% (see
section 2.1).

The NBS fission source is not generally used for low-energy integral
detectors because of uncertainties in the graphite return background which has
a near-1/E distribution of neutron fluence. The graphite return correction
for the 235uU(n,f) reaction, for example, is nearly 10% and self-shielding of
the cavity return fluence by the fission disks and by the 235y activation foil
makes this correction uncertain to a factor of about one third. The one meter
cavity arrangement at the SCK/CEN is better suited for low-energy response
detectors [9].


Appendix A

DETECTOR ASSEMBLY WORKSHEET FOR CAVITY FISSION SOURCE

Assembly Date: 1/17/86 Irradiation 8601 EOI: 21:37:00 Jan 20, 1986
SOI:11:01:00 EST, Jan 18, 1986
(1) (2)
Assembly Piece Depth Measurement S (mils) Zz
(239-S)
Ident- thk. piece sum micro- assigned (D= 443)
ification meter = 324H19
239 Z2=S -120
Spacer 70
Au-1L 5 167 72(Au-1) -48
Spacer 10
Ni-3 lL 148 91(Ni-3) -29
143 142.5
Spacer 11
Ni-4 11 126 113(Ni-4) -7
121
Spacer 11
Ni-5 i 105 134 (ni-5) +15
99 101
Spacer 11
Ni-6 1 83 159(Ni+6) +40
77 80
Spacer 15
s-10 6 59 180(Sc) +61
Au~2 5 53 186 (Au) +67
51 55
Spacer 60
T=249
Assembled Cd. Capsule: 324 = f

Cd. Capsule Holder: 367


NBS MEASUREMENT SERVICES:
ACTIVATION FOIL IRRADIATION BY
REACTOR CAVITY FISSION SOURCES

George P. Lamaze

James A. Grundl

Center for Radiation Research
National Measurement Laboratory
National Bureau of Standards
Gaithersburg, MD 20899

U.S. DEPARTMENT OF COMMERCE, C. William Verity, Secretary
NATIONAL BUREAU OF STANDARDS, Ernest Ambler, Director

Issued March 1988


aon

uw
&

A
A

55

ay
wo

Appendix F

REM SAT260 ...11/21/83

! THE DATA VARIABLES IN

' SEQUENCE:

! X1(1)..-FIRST COUNTER
! c(t) ~-Cu

t X2(1)...Cac

! Té1) ~seck

' XS(1)...DAY

‘ X6(1)...hhmmss

! X7(1)...TIME INTERVAL

REM Fis IS THE DATA FILE

REM NAME

PRINTER IS 701,80

PRINT CHRS (27) &"&11L"

CLEAR

DIM C(260), X14 (260) , X5(260) ,X3 (260) ,X7 (260) X46 (260), T6260)
DISF “DATA FILE TO BE USED:"

INPUT Fis

ASSIGN# 1 TO F1s

READ# 1,1 ; N

FOR I=1 TON

READ# 1,141 ¢ X1¢1),C(T),X3(1), TCT) XSC1)  X6C1)  X7 ET)
NEXT I

ASSIGN# 1 TO *

TCL), X11), 001), X301),X7 (1) =0

PRINT "SATZ60 DATA ANALYSIS"

PRINT " 11/83 version"

PRINT

PRINT "N = "sN

PRINT

PRINT “DATA VERIFICATION FOR “;Fisé

PRINT " cl tu tgc Ahmamss SeCCkS2

FOR I=i1 TON

PRINT USING 340 ; I,X1(1),C(1),X3¢(1),X6(1), TCT) X7 CT)
IMAGE 3D,2¥%,7D,3X, 7D,3X,7D,5X, 6D, 3%, 46D, 3X, 4D

NEXT I

FRINT

FOR I=1 TON

NEXT I

PRINT

PRINT

A1,A0=X5¢1)

FOR I=2 TON

IF XS(T)<>A1 THEN A1l=X5¢1)

NEXT I

FRINT USING 490 ; AQ

IMAGE "DATA GOES FROM DAY", 4D

PRINT USING S19 ; Al
IMAGE " TO DAY",4D
PRINT

DISF

© PRINT “DIFINING THE IRRADIATION FERIOD"
© DISF “DEFINING THE IRRADIATION PERIOD"
© DISF

O PRINT

580 DISF “WITH WHICH INTERVAL WOULD YOU"

39
60
61

© DISF “LIKE ANALYSIS TO START "3
O INFUT ZS
O IF Z5*1 THEN 630

F-1

HFsecTM


The fission neutron fluence is perturbed by a small cavity return contri-
bution, and also by scattering in fission source disks, in encapsulations, and
in the detector foils. The correction for these effects, based on calcula-
tion, is given as the fractional departure for each detector (use) from its
free-field activity:

total :
free-field
measured = aad {1 + ue]
activity activity sc

This correction depends markedly upon reaction threshold because of the change
in neutron energy caused by scattering in the nearby materials of the
encapsulation. Components of Use > obtained from detailed Monte-Carlo

calculations [15] of the source-detector capsule with a representative foil
stack, are as follows:

change in changes in threshold detector response
neutron for the indicated changes in neutron fluences
fluence

Egg 7 162 MeV Egg = 2.1 NEV Eg, = 6.5 MeV

cavity return 41.2% +(0.17+ 0.1)%  < 0.1% < 0.1%
encapsulations 42.5% +(0.5 + 0.3)% -(0.2 + 0.3)% -(1.3 + 0.3)%
(incl. source
disks and cd box)

foil stack +1.5% +(0.0 + 0.4)% -(0.6 + 0.4)% -(0.8 + 0.4)%

Uge (8): +(0.7 £ 0.5)% -(0.8 # 0.5)% -(2.1 + 0.5)%

The uncertainties listed are lo. The spread of the foil-stack component of
Use iS associated with foil position, It is possible to derive from the Monte

Carlo calculation individual foil-to-foil scattering factors as a function of
foil separation. These correction functions can be applied to a range of foil
stack arrangements beyond the one calculated as long as the foil-to-foil
scattering is mainly first collision, An overall uncertainty of + 0.5% (1c)
is assigned to gee This uncertainty has been estimated from the distribution

of the scattering corrections from a large number of foil positions in several
different Monte Carlo calculations.

It is the small source-to-detector distance which makes it possible to
obtain nearly pure fission neutron fluences in small cavities. The cavity
return fluence at the cavity center is about 1.2% of the uncollided fission
neutron fluence. This return fluence gives rise to threshold detector
responses that are between 0.05% and 0.3% of the free-field fission neutron
response depending upon the reaction threshold.


2. NEUTRON FIELD CHARACTERIZATION

The energy spectra of fission neutrons are very similar for all fission-
able nuclides and both test and power reactors exhibit fission-spectrum-like
components in the MeV energy range. The 235U fission neutron spectrum may be
described in terms of a broad energy range,

lower bound median upper bound
Ep 299) Ep(p=0.5) E,(p=0.01)
235y 0.08 MeV 1.57 MeV > 7.2 MeV

where p = fraction of neutron spectrum above Eps and a coarse seven-group flux
display, $(E)AE, as follows:

E limits: 0 0.25 0.8 1.5 2.3 3.7 8 12 MeV
o(E)dE: 0.054 0.197 0.229 0.195 0.192 0.127 0.006

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

Although the neutron spectrum listed is specifically for thermal-neutron-
induced 235U fission, the fission spectrum for fast-neutron-induced fission is
little different. The change in median energy is about 0.05 MeV.

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

The traceability path of NBS neutron field strengths begins with NBS-I,
the U.S. national fast neutron source standard [4]. This source consists of a
beryllium sphere, 4.0 cm in diameter, at the center of which is placed one
gram of radium in the form of RaBry. The radium is enclosed in platinum-
iridium capsules of 0.2 mm wall thickness. This source has been calibrated by
several independent methods [11-13] and the source strength (n/s into 41
steradians) is now established as 1.243 x 106 + 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 [14]. The uncertainty of the
source strength obtained in this fashion is + 1.1% (lo). The 235U fission
neutron fluence in the cavity is established by means of fluence transfer from
the 252Cf fission neutron irradiation facility. The neutron fluence at the
midplane of the source-detector capsule for a source disk separation of 12 mm
is about 0.12 n/cm2 for each neutron emitted from the fission disks which
corresponds to a fluence rate of ~ 3.5 x 1019 n/cm2 s with the reactor at
20 megawatts full power. The fission neutron fluence rate is minimum at the
midplane between the two source disks. At 1 mm above or below the midplane,
the mean fluence for a 12 mm diameter disk detector is ~ 4% higher than at the
midplane, Radial variations are more severe: center-to-edge fluence ratios
are 1.2 for 12 mm diameter disks. This gradient does not require that
detector disks be extremely uniform. A radial mass nonuniformity of 20%, for
example, distributed linearly in a detector foil changes the activation by
less than 0.5% compared to a uniform mass distribution. Also, the nickel
fluence monitor included in each irradiation package yields a fluence averaged
over the radial variation.


aaa

aaa

aaaa

BL=BL/FLOAT (NTL)
BLOW(I)=BL
BH=0.0
DO 66 JK=1,NR(I)
66 BH=BH+Y(IPLUS+JK-1)
NTR=NR(I)
XH=FLOAT (IPLUS)+(FLOAT(NTR) /2.0)
BH=BH/FLOAT (NTR)
BHIGH(I)=BH

BACKGROUND EQUATION

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

BKG(1I)=B

TOTAL(I)=T

PEAK ENERGY
EGY (1)=A+Z*CPK+C*CPK*CPK
PERCENT ERROR - SQRT (T+((U-L)/2)*(BL+BH))/N

CONST=(DD-CC) /2.

SUM=BL+BH

ER(1)=SQRT(T+CONST*SUM) /YAA

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

BINT=YAA*CLTM/TIME

RATEUN=BINT/CLIM !DEADTIME CORRECTED RATE

RATESC IS RATE CORRECTED FOR WITHIN COUNTING TIME DECAY
FOR VERY SMALL CORRECTIONS, THIS CORRECTION IS NOT MADE

IF(.NOT. PULSE)GO TO 74
RATESC=BINT/CLTM
RATETO=RATESC

PLSRAT=RATESC ! ANALYZER PULSER COUNT
PULSE = .FALSE. ! (PLSRTE/PLSRAT) HAS BEEN DETERMINED FOR ST 86
CORR=PLSRTE/PLSRAT
GO TO 90
74 CORR=1.

IF(CLTM*ALAM .LT. .001)GO TO 87
RATESC=BINT*ALAM/ (1-EXP(-ALAM*CLTM) )
GO TO 86
87 RATESC=BINT/CLTM
86 RATETO=CASRAT*RATESC*CORR*EXP (ALAM*SECS)
90 TYPE 81,1,IC(I),ID(I),EGY(I),B,YAA,ER(I),RATEUN
WRITE(4,81)1,IC(I),ID(1),EGY(1),B,YAA,ER(I),RATEUN
81 FORMAT(1X,14,14,15,F7.1,F10.1,F13.1,F7.2,1X,1PE12.5)
IF (SECS .LT. 1.) GO TO 40
TYPE 89,RATESC,RATETO
WRITE(4,89)RATESC ,RATETO
89 FORMAT (/1X,'DECAY CORRECTED RATE = ',F10.3,' COUNTING
& RATE AT TO = ',1PE12.5/)

D-14


Qaaaaa

c

4 FORMAT(I1)

5 FORMAT('.P') ! THIS IS USED BY PGM EDPAGE
TYPE 8

8 FORMAT(/" LOCK SHIFT KEY'/)
CALL PARAMS(TEM)

10 CALL GETFIL(5, IFILI, IFILO)

FORMERLY SUBROUTINE READFI

READ AND PRINT TITLE AND COLUMN HEADS AND INPUT SPECTRUM DATA
FROM THE INPUT FILE. IT IS ONLY CALLED BY THE MAIN PROGRAM.

READ(3'1)(IWORD(I),I=1,34)
READ(3"IWORD(5))ADATA
PRINT *,ADATA
READ(ADATA,7 )ADATA(:16),NDAY, AMONTH,NYR,IHR,ADUM,MINIT,ADUM,
1NSEC , TIME, CLTM
7 FORMAT(A16,12,A5,12,15,A1,12,A1,12,2F11.0)
PRINT 7,ADATA(:16),NDAY, AMONTH,NYR, IHR,ADUM, MINIT ,ADUM,NSEC,
1TIME,CLTM
WRITE(4,7)ADATA(:16) ,NDAY, AMONTH,NYR, IHR, ADUM,MINIT, ADUM,NSEC,
1 TIME, CLTM
SE SEC
IYR=1900+NYR
NLP=IMOD(NYR,4)
IF (AMONTH. EQ. '-JAN-') IMON=
IF (AMONTH. EQ. '-FEB-' ) IMON=2
IF(AMONTH. EQ. '-MAR-' ) IMON=3
IF (AMONTH.EQ. '-APR-") IMON=4
IF (AMONTH.EQ. '-MAY-") IMON=5
IF(AMONTH. EQ. '-JUN-' ) IMON=6
IF(AMONTH.EQ. '-JUL-") IMON=7
IF (AMONTH. EQ. '-AUG-' ) IMON=8
IF (AMONTH. EQ. '-SEP-' ) IMON=9
IF(AMONTH. EQ. '-OCT-' ) IMON=10
IF(AMONTH.EQ. '-NOV-")IMON=11
IF(AMONTH. EQ. '~DEC-' ) IMON=12
IDAY=(IMON-1)*30+NDAY
IF(IMON.EQ.2.OR.IMON.EQ.6.OR.IMON.EQ.7)IDAY=IDAY+
IF(IMON.EQ.3)IDAY=IDAY-1
IF(IMON.EQ.8 )IDAY=IDAY+2
IF(IMON.EQ.9.OR.IMON.EQ.10)IDAY=IDAY+3
IF(IMON.EQ.11.OR.IMON.EQ.12)1DAY=IDAY+4
IF LEAP YEAR(NLP=0)ADD A DAY AFTER FEBRUARY
IF(NLP.EQ.0.AND.IMON.GE.3)IDAY=IDAY+1
READ(3"IWORD(6))TITL1
READ(3' IWORD(7) )TITL2
PRINT *,TITL1 ,TITL2
WRITE(4,40)TITL1 ,TITL2
4O FORMAT (4(1X,A64/) )
INC=0
DO 20 IN=IWORD(31) , IWORD(31) +255,
READ(3'IN)(¥(I),I=INC+1 ,INC+32)
20 INC=INC+32

ICHAN=INC



DISF "WITH WHICH INTERVAL DO YOU WANT "

DISF “ANALYSIS TO CONCLUDE";

INFUT N

T(OS) .X1(05),C (QS), X35 (5) ,X7 (QS) =O

ZO=N~ZS+1 @ REM ZQ IS CORRECT DIVISOR FOR AVERAGES INSTEAD OF N, WHEN ZO I
1

PRINT “ANALYSIS INCLUDES INTERVAL: "3;Z5

PRINT "THROUGH INTERVAL: "iN

PRINT

PRINT "E.S.T. AT S.O.1. 2 ":X6(Z5-1)

BO PRINT "E.S.T. AT E.O.T. 2 "3X6(N)

PRINT

X1=N

T

Tis0

C1,S1=0

790 FOR Q5 TON

B00 T(I)=T(I)/32

B10 Ti=Ti+T (I)

20 $1=Si+C (1)

B30 Z1=X6(1) k10"~(-4)

B40 X6(T)=IF(Z1) k5600

850 Z1=FF(Z1)*k100

860 X6(1I)=X6( 1) +IP(Z1) KGO+FP(Z1) K100

870 X(T 6(1)+X5(1) #86400

880 IF I=@5 THEN 900

B90 C1=C1i+C(1)/T CL) k(XH(1)-X6(I-1))

900 NEXT I

910 T=X6(N)-X6 (5)

915 PRINT "IRRADIATION TIME :"3T3"s"

916 PRINT

920 C1=C1/T

2180 FRINT

2190 PRINT "THE CORRECTION IS THE RATIO OF“
2200 PRINT "ACTUAL ACTIVITY AT EOI TO THAT"
2210 PRINT "WHICH WOULD HAVE BEEN PRODUCED"
2220 PRINT "BY THE SAME FLUENCE AT CONSTANT"
2230 PRINT "POWER" @ PRINT

2240 CLEAR

2250 DISPF "WHICH LAMBDA DO YOU WANT?"

2260 DISF " NUCLIDE t1/2
2270 DISP "1. In 4. ABSH «2926 -5"
2280 DISF "2. Sc-46 83.85d 9.568E-8"

2290 DISP

Ni 70.81d

z0E-7"

2300 DISP Mn-S6 2.576h 7.4744E-5"
2310 DISF Mn-54 312.5d 2.567E-8"
2320 DISF "6. Au 2.696d 2.9757E-6"
2330 DISF "7. Na 15.00h 1.2836E-5"

2340 DISF "8. Sc-47 3. 40d 2.3596E-6"

2350 DISF "9. Sc-48 43.8h 4.3959E-6"

2355 DISF "10. Ba-La 12.789d 6.2373E-7"

2360 INPUT A

2370 DISF "WAIT"

2380 ON A GOTO 2390, 2400, 2410, 2420, 2430, 2440, 2450, 2460, 2470, 2475

2390 L=.00004292 & “In" @ GOTO 2480

2400 L=.90000009568 © Lt="Gc 46" © GOTO 2480

2410 L=,.0000001133 @ L$="Ni" © GOTO 2480

2420 L=.000074744 & L%="Mn-46" ® GOTO 2480
F-2

r



Appendix D

(7G IIIT IE a TR
STP2D8.KIN

MODIFIED 10/1/84 TO MAKE VARIABLE Y 'VIRTUAL' FOR 8000 CHANNELS
FORMERLY STPBG.FOR FROM PDP-11 RT-11 OPERATING SYS.

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

MODIFIED 9/25/82 BY MCGARRY (HAD TO CHANGE READ STATE IN READFI)
BACKGROUND IS COMPUTED FROM AVERAGES OVER REGIONS TO THE LEFT AND
TO THE RIGHT OF THE PEAK REGION, RESPECTIVELY. THE BACKGROUND
REGIONS NEED NOT BUTT UP AGAINST THE PEAK AREA.

THIS PROGRAM WILL DO AN ENERGY CALIBRATION, LOCATE PEAKS,

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

THE DATA OF THE SPECTRA MUST BE STORED ON A FILE WITH ANY STANDARD
ORTEC .SPC FILE SPECIFICATION. PEAK AND END OF IRRAD~

IATION PARAMETERS ARE ENTERED IN SUBROUTINE PARAMS AND ARE READ
FROM FILE LIMITS.DOC OR FROM THE KEYBOARD OR JUST NOT USED. WHEN
THEY ARE NOT USED, THE PROGRAM WILL FIND THE PEAK PARAMETERS AND
OUTPUT THEM ON FILE LIMITS.PKS. IF THEY ARE INPUT FROM THE KEYBOARD
THEY WILL ALSO BE OUTPUT TO LIMITS.TEM. THEY SHOULD BE CONCATENATED
TO LIMITS.DOC, IF ONE WISHES TO SAVE THEM. THE PROGRAM WILL STORE
THE INTERCEPT, SLOPE AND SECOND ORDER CONSTANT FOR THE ENERGY CALI-
BRATION CURVE ON KEV.DOC FROM THE ENERGY CALIBRATION ROUTINE OR USE
AN EXISTING FILE WITH THAT DATA. THE RESULTS ARE WRITTEN ON THE FILE~
NAME.RES AS WELL AS ECHOED TO THE TERMINAL IN SUBROUTINE EPEAK.

aAAgMFaAaaAaAAAAAAAAAAaAAaaAaaaad

(RHE III III ET DI IR IER
(33 III IG GGG Iga IID IEE

Cc

c THE MAIN PROGRAM CALLS SUBROUTINES PARAMS, GETFIL,
Cc READFI, CAL, LOCATE, AND EPEAK IN THAT ORDER.
Cc

OETTETETTCLETreiTrerrretiretiritirtertttrittrtttrreti ttt ttt ttt irri
ETTTLILT LETT TET EILEL TT LTLELET Te LtLer rset i iiit titi itt LireeeLLLLLELLeiii
c
DIMENSION IWL(40),IWH(40),PK(40),
&IC(4O),ID(40) ,NL(4O) ,NR(HO)
CHARACTER IFILI(20),IFILO(20),REP
LOGICAL TEM
DIMENSION IWORD(34)
CHARACTER*128 TITL1,TITL2,ADATA
CHARACTER*5 AMONTH
CHARACTER*1 ADUM
DOUBLE PRECISION XME(4)
VIRTUAL Y¥(8192)
LOGICAL*1 PRMS
COMMON / ADCAM/ TADCAM
COMMON /NCHAN/ ICHAN
COMMON /PKLIM/ NUM, IWL,IWH,IC,ID,NL,NR,PK,IBW,SIG
COMMON /FILNMS/ IFILI,IFILO
COMMON /IPUT/XME
COMMON/LAMBD/IYRO, 1DAYO , THRO ,MINITO,SECO, ALAM, CASRAT,
&IYR,IDAY, IHR,MINIT , SEC, PRMS
COMMON /SPEC/TIME,CLTM,PLSRTE
2 FORMAT(A1)
3 FORMAT(I4)


1. DESCRIPTION OF SERVICE

The National Bureau of Standards (NBS) maintains standard fission neutron
fields (or benchmark fields) to supply calibration irradiations and certified
fluence standards. The NBS Cavity Fission Neutron Source provides a 235y
thermal-neutron-induced fission neutron field spectrum. Known fluences are
based upon the known source strength for the National Standard Ra-Be Neutron
Source, NBS-I [1]. The NBS calibration service 44090C [2] (formerly 8.1R)
operates in the following way: (i) dosimetry sensors (metal foils, nuclear
track detectors, wires, crystals, etc., supplied by the customer or by NBS)
are irradiated to a measured neutron fluence in a fission-spectrum neutron
field (for short lived reaction products, the samples are irradiated to near
saturation and the fluence rate specified), and (ii) the irradiated dosimetry
sensors are shipped to the customer, followed by a test report stating the
fluences and associated uncertainties. Typical maximum fluences are of the
order of 3 x 1015 neutrons/cm2, with combined (lo) uncertainties of + 2%.
Dosimetry sensors must fit within a cylindrical cadmium enclosure of
20-mm-inside diameter and 6.5-mm-inside height.

For a typical activation detector, the customer measures the activity of
the irradiated sensor with his own detector system and corrects the activity
to obtain the activity at the end of the irradiation (EOI). This activity is
then corrected for scattering effects (see section 2.1) to obtain the 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=rA+CHN Geo (1)

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

X = decay constant (s-1) of the reaction product,

C = decay correction factor (dimensionless) given in the test report

(this corrects for activity lost during the irradiation),

N = number of atoms in the target isotope,

& = fission-spectrum-averaged reaction cross section (cm2),
and @ = 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 2.3.
xen = 7° (2)


D.
In practice, Study is simply the time history corrected counting rate

ratio of the sample ikradiated in the unknown field to that irradiated in the
235y 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 study?

6. SAFETY

NRC and NBS regulations require that all personnel working at the NBS
reactor undergo a radiation safety and awareness training course. This course
acquaints the worker with the applicable radiation regulations as well as the
hazards of radiation. Further training of any workers using the thermal
column facilities is also mandatory. Very high radiation doses can be
obtained through carelessness in the vicinity of the thermal column, A
specific hazard for the cavity fission source irradiations is the fission-
product gamma activity from the 235U disks after they have been irradiated,
Radiation doses can be reduced by: (1) waiting as long as possible before
handling the disks, (2) having a well thought out plan of action for handling
the disks, (3) working rapidly, and (4) using shielding whenever possible.
Since the latter two often conflict, personal judgment based on experience
must be used. Item 2 can be facilitated by performing "dry runs" with "cold"
or dummy sources,

The irradiated samples will be radioactive, but at a very low level
(usually a few microcuries). There is therefore little hazard from the
samples, but all radioactive shipping regulations must still be followed. The
samples should be stored in a lead shielded storage area while awaiting
analysis or shipping.

15


c ANSWER NO TO THE GAINSHIFT REQUEST IN SUBROUTINE EPEAK. THEN
c THE PROGRAM WILL NOT ACCESS THIS SUBROUTINE.

VIRTUAL ¥(8192)
COMMON /SPEC/ TIME,CLTM
ISTEP=1
I=I+ISTEP
IF (¥(I).LT.O.)¥(I)=-¥(1)
YL=¥(I-1)+SQRT(Y(I))
YH=¥(1I+1)-SQRT(Y(I))
IF((YL.LT.¥(I)).AND.(¥(I).LT.YH)) GO TO 50
Go TO 100
50 IL=I
52 DIF=¥(1+1)-¥(I)
IF(DIF.LT.O.0) GO TO 51
IsI+1
GO To 52
51 YP=¥(I)
IK=1
53 I=I+1
YL=¥(I-1)+SQRT(¥(I))
YH=Y(I+1)-SQRT(¥(I))
IF((YL.GT.¥(I)).AND.(¥(I).GT.YH)) GO TO 53
54 IH=I
YL=¥(IL)
YH=¥ (IH)
BK=(YL+YH)/2.0
PW=IH-IL
IF(PW.LE.3) GO TO 100
70 CNT=YP-BK
TOT=YP+BK
IF(TOT.LE.0.) GO TO 100
IF(CNT.LT.SIG*SQRT(TOT)) GO TO 100 ! SIG = PEAK SIGNIFICANCE
SXY=0. ! SELECTION CRITERION
SY=0.
ILOW=IK-2
IHI=IK+2
DO 75 K=ILOW,IHI
CK=K
SXY=SXY+CK*Y (K) *¥ (K)
SY=SY¥+¥(K)*¥(K)
75 CONTINUE
CPK=SXY/SY
GO TO 101
100 CPK= -10.0
101 RETURN ! TO SUBROUTINE ELOC, LOCATE, OR EPEAK
END ! END OF SUBROUTINE CHPEAK
CER RK KKK HK KR KKK HHH KR KEK KR KK HEHE KKH K KEKE KKK HR R KKK KEK K KHER RR HER EERE EK
c
SUBROUTINE LOCATE(Y)

SUBROUTINE LOCATE DETERMINES PEAKS AND BOUNDARIES AND
PASSES INFORMATION TO SUBROUTINE EPEAK VIA COMMON BLOCKS.
IT WRITES ON THE OUTPUT FILE FILENAME.RES.

aaaa

D-9


DIMENSION EGY(40),IPK(40),IWL(40),IWH(40),PK(40),
&IC(4O), ID(40), NL(40) ,NR(4O)
VIRTUAL ¥(8192)
COMMON/ ADCAM/T.ADCAM
COMMON /NCHAN/ ICHAN
COMMON /ENEG/ EGY
COMMON /PKLIM/ NUM, IWL,IWH,IC,ID,NL,NR,PK,IBW,SIG
IF(IADCAM.GT.0)GO TO 200
OPEN(UNIT=2, FILE='KEVO.DOC' ,STATUS=" OLD")
Go TO 201
200 OPEN(UNIT=2,FILE='KEV1.DOC' ,STATUS="OLD')
201 READ(2,1)A,Z,C
1 FORMAT(3E17.10)
CLOSE(2)
NUM-O
TYPE 2
2 FORMAT(/'$PEAK SIGNIFICANCE : ')
ACCEPT 3,SIG ! USED AFTER STMT 70 IN SUB CHPEAK
FORMAT (F6.0)
WRITE(4, 4)
WRITE(4 4)
FORMAT (' KRRERRRRRKAA AAA RHRRAAERHAREREREERERERE
RRR RK EK K KEK RHR EK HK ERE ER EK KKK KEK ERE)
WRITE(4,5)
TYPE 5
5 FORMAT(4X,"PEAK LOCATOR'//' PEAK CHAN ENERGY FWIDTH'/)
I=1
10 CALL CHPEAK(CPK,IH,IL,1.SIG.Y)
IF(CPK .LT. 0) GO TO 100
EN=A+Z*CPK+C¥CPK*CPK
NUM=NUM+1
EGY (NUM)=EN
IFW=IH-IL+1
IWL(NUM)=IL
IWH(NUM)-IH
IC(NUM)=IL
ID(NUM)=IH

w

NR(NUM)=IBW

PK(NUM)=CPK

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

TYPE 6,NUM,CPK,EN, IFW

6 FORMAT(15,F8.2,F9.2,3X,15)
100 IF(I.LT.ICHAN) GO TO 10

RETURN ! TO SUBROIUTINE EPEAK

END ! OF SUBROUTINE LOCATE
CHEK KK KK KK EK KEKE K HHH KKK RK KKK RK KK KKK KKH KKK KEK HEHEHE HK ERR EERE RRR KK RR EEK
c

SUBROUTINE EPEAK(Y)

SUBROUTINE EPEAK USES PEAKS AND BOUNDARIES FOUND BY
SUBROUTINE LOCATE OR THOSE FROM SUBROUTINE PARAMS
IT CREATES ALL THE OUTPUT RESULTS WHICH ARE WRITTEN

aaaa

D-10


ASSIGNMENT OF UNCERTAINTIES (+1 o)

e Flus Transfer Measurements from the 252Cf Standard Fission
Neutron Field to the 235 Cavity Fission Source:

Percent
(1) Absolute neutron source strength of the 252Cf source 0.9 S+R
(2) Distance measurement 0.6 R
(3) 252CF to 235y spectrum-averaged cross section ratio
for Pu(n,f) monitor 0.1 Ss
(4) Correction for neutron scattering for Pu(n,f) 0.9 Ss
(5) Reproducibility of flux transfer procedure (including
counting statistics, and electronic system stability
over period of measurement) 0.4 R
Subtotal (summed in quadrature) = 1.5
e@ Preparation of Neutron Fluence Standards at the
235 Cavity Fission Source:
(1) Gradient corrections and positioning 0.4 S+R
(2) Corrections for neutron scattering for 115In(n,n') 1.2 R
(3) Reproducibility of fluence monitoring including random
summing corrections and uncertainties in activation
decay corrections 0.4 R
Subtotal (summed in quadrature) = 1.3
TOTAL UNCERTAINTY = 2.0

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


4, Activation decay rate factors:

T= 2.10960S 5°Co »98Au *6Se

A- 1.1334x10-7 2.9762x10-© 9.5700x10-®
satfac 2.363x10-? 4 .663x10-* 1.999x10-?
G= satfac/T 1.1200x10-7 2.210x10-© 9.47 4x10-%
C= G/A = +9881 -TH26 +9900

Note: For a constant fluence rate irradiation, satfac = (1-e 7)

5. Fission Neutron Fluence at Midplane (z = 0): 4

°

o(x

ef)

9 = Kop * a(Xy5) ” [p/a)
counter response of fluence monitors

corrected for saturation (SATFAC code)

calculated cross section ratio for X_. to Xo5,
fluence transfer reaction (before 9-83: In(njn) 5

after 9-83:this factor is included in Kor)

calibration factor (i.e. composite of detector
efficiency parameters) (For GEM detector, single shelf,
zero thickness nickel foil, Kor = 1.711 x 10°)

= s ,_5385__ 1s
5 = 1ATI1KI08 7-73 anys 8 .226x10

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

isd. Z 4(z)/(0) free-field fluence u_, activity by Total
__ mils @ foil pos. scattered flux(%) Error
Ni-3 -29 1.0178 8.373x107® -.7
Ni-4 -T 1.0010 8.235x10?§ 7
Ni-5 +15 1.0048 8.266x10'5 -.7
Ni-6 +h0 1.0339 8.506x1075 -.7
Au-1 ~48 1.0488 8.628x1075 +4
Au-2 +67 1.0952 9.010x101§ +65
S-10 +61 1.0789 8.876x10!® +.6


ABSTRACT

This document describes the National Bureau of Standards (NBS)
calibration service 44090C (formerly 8.1R) which operates in the
following way: (i) dosimetry sensors (metal foils, nuclear track
detectors, wires, crystals, etc., supplied by the customer or by
NBS) are irradiated to a measured neutron fluence in a fission-
spectrum neutron field (for short lived reaction products, the
samples are irradiated to near saturation and the fluence rate
specified); and (ii) the irradiated dosimetry sensors are shipped
to the customer, followed by a test report stating the fluences or
fluence rates and associated uncertainties. The scope and
philosophy of the service, the neutron field characteristics,
irradiation procedures, and the uncertainties in the reported
neutron fluences or fluence rates are discussed. Typical maximum
fluences are of the order of 3 x 1015 neutrons/cm2, with combined
(1c) uncertainties of + 2%.

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

iv


Y=-,0017+(..00356*EXP(-.0815*X) )
Y=Y*TH(J)/10.
SCAT=SCAT+Y
550 CONTINUE
ALSCAT=SCAT~.002+(.0017*TH(I)/10.)
Go TO 800
600 DO 650 J=1,N
IF(I.EQ.J)GO TO 650
X-ABS(POS(J)-POS(I))
=-.002+(.002*EXP(-.147*X))
Y=Y*TH(J)/10.
SCAT=SCAT+Y
650 CONTINUE
ALSCAT=SCAT-.011
GO TO 800
800 WRITE(7,102)ANAME(I),GRAD(1),IT(1),POS(1),TH(I),SCAT,ALSCAT
WRITE(3 , 102) ANAME(I) ,GRAD(I),IT(),POS(I),TH(I) ,SCAT, ALSCAT
102 FORMAT(1X,A4,F10.4,110,2F10.1,3X,2E12.3)
1000 CONTINUE
CLOSE(3)
IF (NF.NE.5)CLOSE(NF)
STOP
END

E-2


40 CONTINUE
TYPE 9 ! LINE FEED
TYPE 3
WRITE(4,3)
WRITE (4,50)
50 FORMAT (/,15X,'DETAILS REGARDING THE BACKGROUND SUBTRACTION’ /)
WRITE (4,55)
WRITE (4,56)

55 FORMAT(2X,'PEAK AREA LOW BKG HIGH BKG"
&,! BLOW BHIGH BL/BH  BKG/T % ERROR')
56 FORMAT(2X,* LIMITS LIMITS —_ LIMITS")

WRITE (4,9)
57 FORMAT(2X,12,1X,14,'-",14,1X,T4,'-",14,1X,14,'~",14,
&1X,F7.1,1X,F7.1,1X,F7.3,1X,F7.4,1X,F7.2)
DO 58 I=1,NUM
D WRITE (4,59) CPK,IC(I),ID(I),NL(I),NR(Z)
D59 FORMAT (5X,F9.2,14,14,12,12)
11=IWL(1)-NL(I)+1
I2=IWH(I)+NR(I)-1
21=BLOW(1) /BHIGH(I)
Z2=BKG(1)/TOTAL(I)
58 WRITE(4,57)1,1C(1),ID(I),11,IWL(1),IWH(I),12,BLOW(I),
&BHIGH(I),Z1,Z2,ER(I)

WRITE (4,9) ! LINE FEED
WRITE (4,3)
WRITE(4 ,3)
52 WRITE(4,9)

TYPE 9 ! LINE FEED
RETURN ! TO MAIN PROGRAM
END ! OF SUBROUTINE EPEAK

c

c

(HOGG EGE II GGG HEI GI IER I
c

D-15


*winagoads uoujnau UOLSSLS 41 ay $0 uOLzeNLeAS SGN aUL ‘Zz ‘BL4

GES

¢ CAPW) ADYANA NOYLNAN

1 Z1 6 g € ft)
ee ME Ht | at a
qt
: | \ f
| q
_
= | v
E | 8
z H ! m
+ | Vv
£ #9
: | zt
— _ | ne x. t
E | z
)
a 7
Ht ee tS

WNYLISdS NOULNAN NOISSIS SEeN


The expression which governs this fluence transfer is

$ = Pow , Sct,
cvy op - cf

cf Covy

See . Seg (PED) YogPED) . Povy -3

Bey Seyy(?Ep) Vey (PE) cg

cf

where ® = fission neutron fluence and “cvy" denotes the 235U cavity fission
source and "cf" the 252Cf fission neutron calibration fluence,

R

the reaction probability, and
ED = truncation energy (p = 0.95 commonly).

The ratios on the right side of the equation involve decay correction
factors (G), truncated reaction cross sections o(>E,), spectrum fractions
v(>E,), and the ratio of the monitor foil activity, D, in the cavity to that
at the 252Cf irradiation facility. Details of the fluence transfer procedure
are given in Refs. [4], [22]. These references also describe a set of
validation experiments, carried out at the SCK/CEN Facility, in which
239Pu(n,f), with negligible fission spectrum shape sensitivity, was
effectively substituted for 58Ni(n,p) and 415In(n,n') as the fluence transfer
reaction. The fluence transfer procedure yields a free-field 235U cavity
fission neutron fluence to an accuracy of + 2.0% (lo).

Fission neutron fluences of ~ 3 x 1015 n/cm2 may be obtained at the NBS
facility in 24 hour long irradiations. The small irradiation volume restricts
multiple foil packages to a diameter of 12 mm and a thickness of a few milli-
meters, The gamma-ray dose from neutron capture in graphite (~ 0.7 kGy/h) and
cadmium (~ 4.4 kGy/h) is about 46% of the neutron and gamma-ray dose from the
fission sources.

4. DETAILED OPERATING PROCEDURES

4.1. ASSEMBLY AND IRRADIATION OF SENSORS

4.1.1. If the sensors are unmarked, they should be marked with a
permanent identifier at this time, preferably by scribing. Sensors should be
cleaned with acetone and weighed after drying. The weighing accuracy should
be 0.2% or better. Masses should be recorded on the worksheet and the plastic
bag containing the sensor. The masses of the nickel and/or indium foils are
necessary to calculate the fluence; other masses are measured to obtain a
complete record of the radiation.

4.1.2. The sensors should be mounted into the detector capsule. The
positions of the sensors should be recorded on the worksheet (see appendix A)
and in the notebook.

10


PREFACE

The calibration and related measurement services of the National Bureau
of Standards are intended to assist the makers and users of precision
measuring instruments in achieving the highest possible levels of
accuracy, quality, and productivity. NBS offers over 300 different
calibration, special test, and measurement assurance services. These
services allow customers to directly link their measurement systems to
measurement systems and standards maintained by NBS. These services
are offered to the public and private organizations alike. They are
described in NBS Special Publication (SP) 250, NBS Calibration Services
Users Guide.

The Users Guide is being supplemented by a number of special
publications (designated as the "SP 250 Series") that provide a
detailed description of the important features of specific NBS
calibration services. These documents provide a description of the:
(1) specifications for the service; (2) design philosophy and theory;
(3) NBS measurement system; (4) NBS operational procedures; (5)
assessment of measurement uncertainty including random and systematic
errors and an error budget; and (6) internal quality control procedures
used by NBS. These documents will present more detail than can be
given in an NBS calibration report, or than is generally allowed in
articles in scientific journals. In the past NBS has published such
information in a variety of ways. This series will help make this type
of information more readily available to the user.

This document (SP 250-14), NBS Measurements Services: Activation Foil
Irradiation by Reactor Cavity Fission Sources, by G. P. Lamaze and

Jd. A. Grundl, is the fourteenth to be published in this new series of
special publications. The scope of the service, the neutron field
characteristics, the irradiation procedures, and the uncertainties in
the reported neutron fluences or fluence rates are described (see test
number 44090C in the SP 250 Users Guide). Inquiries concerning the
technical content of this document or the specifications for these
services should be directed to the authors or one of the technical
contacts cited in SP 250.

The Center for Radiation Research (CRR) is in the process of publishing
21 documents in this SP 250 series, covering all of the calibration
services offered by CRR. A complete listing of these documents can be
found inside the back cover.

NBS would welcome suggestions on how publications such as these might
be made more useful. Suggestions are also welcome concerning the need
for new calibration services, special tests, and measurement assurance
programs .

Joe D. Simmons Chris E. Kuyatt
Acting Chief Director
Measurement Services Center for Radiation Research

iii


OPEN( UNIT=2, FILE="KEVO.DOC' ,STATUS='OLD')
GO TO 201
200 OPEN(UNIT=2,FILE="KEV1.DOC',STATUS="OLD')
201 WRITE(2,50)A,B,C
CLOSE(2)
50 FORMAT(3E17.10)
TYPE 55
55 FORMAT(/' FILE KEV UPDATED. '/)
RETURN ! TO THE MAIN PROGRAM
END ! OF SUBROUTINE CAL
CER HR HHH RHR HHH KKH KKH KKK HHH KH HHH KR RRR K RRR HR ER HR RHR HEHE HHH KR KER KHER K
c
SUBROUTINE ELOC(Y)

IT RETURNS PEAK LOCATIONS WITH THE ENERGY SUPPLIED BY THE
TERMINAL FOR THE ENERGY CALIBRATION, WHICH IS DONE BY CAL.
IT CALLS SUBROUTINE CHPEAK.

aaaa

DIMENSION X(40),E(40)
VIRTUAL ¥(8192)
COMMON /NCHAN/ ICHAN
COMMON /PEAK/X,E,NUM
NUM=0
TYPE 1
1 FORMAT(/'$PEAK SIGNIFICANCE : ')
ACCEPT 2,SIG ! USED AFTER ST 70 IN SUB CHPEAK
2 FORMAT(F6.0)
TYPE 3
3 FORMAT(//4X,"PEAK LOCATOR'//' CHAN ENERGY '/)
I-1
10 CALL CHPEAK(CPK,IH,IL,1,SIG,Y)
IF(CPK .LT. 0) GO TO 100
NUM=NUM+1
X(NUM)=CPK
TYPE 4,CPK
4 FORMAT(1H ,$F7.1,2X/)
ACCEPT 5,E(NUM)
5 FORMAT(F10.0)
IF(E(NUM).LT.1.0) NUM=NUM-
100 IF(I.LT.ICHAN) GO TO 10
RETURN ! TO SUBROUTINE CAL
END 1 OF SUBROUTINE ELOC

CER RRR KHER KKH KK KKK KKK KK IKK KKK KKK RK KEK RK RK RRR EERE RK ERE
Cc
SUBROUTINE CHPEAK(CPK,IH,IL,1I,SIG,Y)

THIS SUBROUTINE LOCATES THE PEAK FROM THE INPUT DATA.

IT STARTS AT CHANNEL I AND STEPS FROM CHANNEL TO CHANNEL
LOOKING FOR PEAKS. IT IS CALLED BY SUBROUTINES ELOC, EPEAK
AND LOCATE. A PEAK MUST BE AT LEAST TWO CHANNELS WIDE IN WHICH
CASE THE SECOND CHANNEL COUNT MUST BE LARGER THAN THE FIRST.
THE ALTERNATIVE IS TO ENTER THE REGION OF INTEREST INTO THE

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

a anaaaaa

D-8


 preces are assembled with 1.4. number up

(2) geror does not include uncertainty in midplane position. Position

error for Indium monitor is relative to adjacent foil.
f = dimension of assembled cadmium capsule (0.326" empty)
D = source disk separation: (f + 119)
Z2=S + (192 - D/2)

= detector position measured from midplane

Foil Mass (g)
Ni-3 0.28609
Ni-4 0.28775
Ni-5 0.28564
Ni-6 0.28721
Au-1 0.28298
Au-2 0.29142
Se S-10 0.04824


Library of Congress Catalog Card Number: 88-600513

National Bureau of Standards Special Publication 250-14
Nati. Bur. Stand. (U.S.), Spec. Publ. 250-14, 56 pages (Mar. 1988)
CODEN: XNBSAV

U.S. GOVERNMENT PRINTING OFFICE
WASHINGTON: 1988

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


L="Mn-S4" & GOTO 2480
"Aut & GOTO 2480

" @ GOTG 2480
L¢="Sc-47" @ GOTO 2480
"Sc-48" & GOTO 2480
"Ba-La"

x

2530 S2=S24C (I) /T I) k (1 -EXP (-C) ) /L¥EXP (—(L# (T- (X6 (9) -X6 (05) 77)
S+ 1 C1-EXP (-C) ) /L KEXP (~— (LCT (X66) -X6 (85) 9)

3560 S=LxS2/C1
2570 O=L4sa

25B0 C3=1-EXP(-(LeT))
70 C2=S/CS
2600 PRINT

2610 PRINT "0000 wesc rene nen ener "
2620 PRINT & PRINT

2630 PRINT "NUCLIDE "SL

2649 PRINT "LAMBDA sls" 1/SEL"

2450. PRINT @® PRINT @ PRINT

2660 PRINT USING 2670 ; S

2670 IMAGE “SATFAC = ",D.DDDDe
2680 FRINT USING 2690 ; @

2490 IMAGE "FLAT CASE SFAC = "D.DDDDe
2700 PRINT USING 2710 3 C3

2710 IMAGE “XACT FLAT CASE SFAC = ",D.DDDDe
2720 PRINT USING 2730 3; C2

2730 IMAGE “CORRECTION = ",D.DDDDe

2740 PRINT @ FRINT & PRINT @ PRINT
2750 CLEAR

2760 GOTO 2240

2770 END


aaaaa

IN THE CASE OF A DECAY LIKE BA, WHICH DECAYS TO LA, ONE MUST

INPUT BOTH OF THE HALF LIVES, THE LONGEST FIRST (BA). IT CALCULATES
THE CASCADING RATIO OF BA ACTIVITY TO LA ACTIVITY AT THE END OF
IRRADIATION, (TA-TB)/TA.

DIMENSION IPK(40),IWL(40),IWH(40),PK(40),
&IC(4O), ID(4O) ,NL(4O) ,NR(4O)
DOUBLE PRECISION KEY, TOKEN
CHARACTER REPL
LOGICAL*1 PRMS,TEM
COMMON/LAMBD/TYRO , IDAYO , IHRO ,MINITO,SECO, ALAM, CASRAT,
&IYR, IDAY, IHR, MINIT , SEC, PRMS
COMMON /FILNMS/ IFILI,IFILO
COMMON /NCHAN/ ICHAN
COMMON /PKLIM/ NUM, IWI.,TWH,TC,TD,NL,NR,PK,1IBW,SIG
DATA NUM, TYRO,IYR,IDAYO,IDAY,IHRO,IHR,MINITO,MINIT /9*0/,
&TA /1.0 E13 /, TB,SECO,SEC /3*0.0/,PRMS/.FALSE./,IBW/3/,
&SIG /40./
TEM=.FALSE.
TYPE 1
1 FORMAT ('$DO YOU WANT TO USE INTEGRATION PARAMETERS ? ')
2 FORMAT(1A1)
ACCEPT 2, REPL
IF(REPL.NE.'Y') GO TO 120
PRMS = .TRUE.
TYPE 8
8 FORMAT('$DO YOU WISH TO READ THEM FROM LIMITS.DOC ? ')
ACCEPT 2,REPL
IF (REPL.EQ.'Y') GO TO 100
TYPE 9
9 FORMAT(' THESE PARAMETERS WILL BE WRITTEN ON FILE LIMITS.TEM',
&/' SO THAT YOU CAN LATER ADD THEM TO FILE LIMITS.DOC')
OPEN (UNIT=2,FILE='LIMITS.TEM' ,STATUS="NEW' )
TYPE 101
ACCEPT 102, KEY
WRITE(2,102) KEY
WRITE(5,122) KEY
TEM = .TRUE.
TYPE 3
3 FORMAT(' GIVE YEAR,DAY,HR,MIN,SEC FOR THE END OF IRRADIATION '
&/' IN 4110, 1F10.0 FORMAT, SEPARATED BY COMMAS '/)
ACCEPT 4,1YRO, 1DAYO, IHRO,MINITO, SECO
TYPE 4,1YRO,IDAYO,IHRO,MINITO,SECO
4 FORMAT(4110,F10.0)
TYPE 5
5 FORMAT(/' GIVE THE HALF LIFE OF THE SINGLE DECAY OR THE TWO HALF'
&/' LIVES FOR A CASCADING DECAY LIKE BA-LA, LONGEST FIRST AND IN'
&/' F OR E14.0 FORMAT, SEPARATED BY A COMMA, IF TWO '/)
ACCEPT 6, TA, TB
TYPE 206, TA, TB
6 FORMAT(2F14.0)
206 FORMAT(1X,2E14.6)
TYPE 7
7 FORMAT(/'$UNITS OF TIME FOR THE FIRST (Y, D, H, M, ORS)? ')


aaa

aaaa

IF(.NOT. PRMS)GO TO 15
IF(.NOT. GAIN)GO TO 15
IF((PK(1).LT.IWL(I)).OR.(PK(I).GT.IWH(I)))TYPE 16
16 FORMAT(' INPUT PEAK IS OUT OF BOUNDS ')
LL=IWL(1)-NL(1)
ILL=LL
LH=IWH(I)+NR(I)
17 CALL CHPEAK(CPK,IH,IL,ILL,SIG, Y)
IF(CPK.GT.5.0)GO TO 14
IF(ILL.LT.LH)GO TO 17
TYPE 13,LL,LH,SIG
13 FORMAT(* NO PEAK WAS FOUND DETWEEN CIIANNEL',I5," AND',I5,
&/' WITH PEAK SIGNIFICANCE =',F5.1)
GO TO 40
14 CPKK=CPK+0.5
TPK=PK(I)+0.5
IDIFF = IFIX(CPKK)-IFIX(TPK)
PK(I)=PK(I)+IDIFF
IWH(I)=IWH(I)+IDIFF
IWL(I)=IWL(I)+IDIFF
IC(1)=IC(1)+IDIFF
ID(I)=ID(I)+IDIFF
IF(IDIFF.NE.O)WRITE(5 ,8)IDIFF
8 FORMAT(' THE PROGRAM PARAMETERS HAVE BEEN SHIFTED',15,
&/' CHANNELS DUE TO A CHANGE IN PEAK POSITION.'/)
15 T=0.0
IMINS=IC(I)
IPLUS=ID(I)
CC=FLOAT(IC(I))
DD=FLOAT(ID(I))
DO 41 J=IMINS, IPLUS
T=T+¥(J)
41 CONTINUE

PEAK LOCATION

SXY=0.
SY=0.
DO 45 K=IMINS,IPLUS
CK=K
SXY¥=SXY+CK*Y(K)*Y(K)
45 SY¥=SY+¥(K)*Y(K)
CPK=SXY/SY

PROGRAM USES NL AND NR CHANNELS ON EACH SIDE TO DETERMINE
BACKGROUND AT PEAK BOUNDARIES. DEFAULT IS IBW.

42 IMINS=IWL(I)
IPLUS=IWH(1)
BL=0.0
DO 64 JK=1,NL(I)
64 BL=BL+Y(IMINS-JK+1)
NTL=NL(I)
XL=FLOAT (IMINS)-(FLOAT(NTL) /2. )


NOTES FOR FLUENCE CALCULATION

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

¢(z) = neutron fluence rate per source neutron at z = distance from midplane
between coaxial disks a distance D apart.

= 0.0470 p7'*16 (1 + 21,2 24)

2 (1 + 1/(21.2 22)!

ag dz
9 Zz

D= 0.0470 pts 16 =

(b) Corrections for scattered neutrons

Monte Carlo calculations from LANL provide net detector activation
associated with collided neutrons:

B-3


2.2. NEUTRON SPECTRUM

2.2.1. Calculation, Fission neutron spectra are most accurately deter-
mined by laboratory measurement because all the calculations have undetermined
fitting parameters. (See, however, Ref. [16] and earlier work cited
therein.)

2.2.2. Measurement [17], [18]. From 1952 to the present, well-
documented measurements of 252Cf and 235U fission neutron spectra have
employed every perfected technique of neutron spectrometry [19]. Accordingly,
fission spectra are the most widely studied fast-neutron energy distributions
in existence. The large body of documented fission spectrum data was
evaluated at NBS in 1975 [18]. The evaluation included an estimate of the
spectrum uncertainties in a multigroup format based on the spread of the
various data sets. The 252Cf and 235U fission spectrum shapes recommended in
the evaluation are based on the sixteen selected measurements listed in
Table 3 of ref. [20].

2.2.3. Evaluated Spectrum. The NBS fission spectrum evaluation has
been chosen for presentation in this document because of the consistency of
its procedures for both 235u and 252Cf fission spectra, and because it
includes an articulated uncertainty estimate [17], [18]. Many other
evaluations are available,

The NBS evaluated spectrum is described up to 20 MeV by means of a
reference Maxwellian, M(E), modified by four piecewise continuous linear
segments below 6 MeV and one exponential segment above 6 MeV. The reference
Maxwellian for the 235U fission spectrum is,

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

and the adjustment functions u(E) are as follows (here 25 denotes 235u):

Energy Interval (MeV) ¥y5(E)
O- 0.25 1+ 0.8E - 0,153
0.25 - 0.8 1 - O.14E + 0.082
0.8 - 1.5 1 + 0.040E - 0.062
1.5 - 6.0 1+ 0.01E - 0.017
6.0 - 20 1.043 exp [-0.06(E - 6.0)/1.043]

The evaluated spectrum is given by x(E) = n(E) + M(E). A 45-group
tabulation of the evaluated 235 fission spectrum is given in table 1 and
plotted in figure 2, The evaluated 252Cf fission neutron spectrum is given
for comparison purposes. Group fluences for other energy group structures may
be derived from this tabulation with the interpolation function recommended in
the footnote to the table. Spectrum uncertainties, given in table 2, are
based on standard deviations from the mean of subsets of measured data from
the final adjusted Maxwellian of the evaluation. This estimate, carried out
in a seven-group structure, is based on the scatter of the data from the
sixteen experiments which were evaluated.


Appendix G

TEST REPORT FOR NEUTRON FLUENCE STANDARDS
PRODUCED BY AN IRRADIATION AT THE NBS

235U CAVITY FISSION SOURCE

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

Sensor Materials Irradiated: Nickel, Gold, and Scandium
(designations: Au-1,2; Sc-S-10; Ni-3,4,5,6)

End of Irradiation (EOI): Jan 20, 1986 at 2137:00 EST (Day 20 of 1986)
REPORT FOR: Ni-5

Free Field Fission Neutron Fluence: @ = 8.266 x 10'* n/cm? + 2 % (1 9)
Physical Description of Material: Mass = 0.28564 g
Form: 12.7 mm dia. X 0.25 mm thk disks
Type: Natural element nickel
Additional Parameters of Irradiation:
Decay Correction Factor: C= 0.9881
Length of Irradiation: T = 210960 s
Departure from free-field activity due to Neutron Scattering (g,)*
We (Ni) = -(0.7 4 0.3) %

Other Radioactivities:
Since the sensors are natural metal samples or in some cases special
alloys, activities from neutron reactions in other isotopes and
impurities may be present. Ordinarily, this will not be a problem

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

The following lines have been observed above the annihilation
energy:
£°Co < 0.1 % of *°Co activity on 1/28/86

This fluence standard prepared by: For the Director
G.P. Lamaze R.S. Caswell, Chief
Ionizing Radiation Division Ionizing Radiation Division


® = free-field fission neutron fluence (n/em*) given in this test
report. (Note: the symbol @ is recommended by ICRU and ASTM
E170.)

reaction probability

ord

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

—— = os <> (3)
CATCIN

{aTcINn = saturated specific activity

LATC] = decay correction factor for average neutron fluence rate. For an

uninterrupted irradiation of length T at constant fluence rate,

{ATC] is equal to ti-e 37],

T = length of irradiation given in the test report.

<$> = average, free-field fission neutron fluence rate (n/om?s) given
in this test report.

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 = YedeCeNe(ord), (4)

where Y is the chain fission yield. When the fission product actvity is
from the daughter of a radioactive parent with a half life longer than
that of the daughter, the appropriate free-field activity at equilibrium
is

A= [A'/(AtHA)]+¥eA+CoNe (0+) (5)

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

daughter.


TABLE 2. 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 (%)
o + AE o* AE
°.0 0.047 13 0.054 16
+ +
0.25 . - . 7
0.8 0.184 +11 0.197 + 4.1
Ls 0.220 + 1.8 0.229 + 3.0
23 0.194 +1.0 0.195 + 3.1
37 0.200 + 2.0 0.192 + 2.0
3 0.146 +21 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.

Of the many alternative descriptions employed for the 235y fission
spectrum, the most common is the Watt-function fit chosen for ENDF/B-V [21] in
dune, 1979:

w(E) = C exp(-E/a) sinh(YbE)
a = 0.988 MeV b = 2.249 Mev-? C = 0.4395

median energy: £,(p=0.5) = 1.64 MeV .

p!
Contemporary time-of-flight data were emphasized in the ENDF/B-V evaluation as
compared with the NBS evaluation which considered all well-documented data.

3. IRRADIATION PROCEDURES:

Disk detector irradiations are monitored with nickel or indium monitor
foils placed between the detectors. The neutron fluence in the detector
capsule is obtained from the measured 58Ni(n,p) or 115In(n,n') activity on the
basis of a fluence transfer calibration carried out at the NBS 252cf
Irradiation Facility. In this procedure, the ratio of the activity obtained
in a known 252Cf fission neutron fluence to that obtained during a cavity
fission source irradiation is translated into a 235U fission neutron fluence
at each monitor foil position in the source-detector capsule.

9