Compendium of

Benchmark Neutron Fields

for Reactor Dosimetiy

Standard Neutron Field Entries

#85-3151

1986 Evaluated and Compiled at the

U.S. Department of Commerce • National Bureau of Standards

NBSIR 85-3151

COMPENDIUM OF

BENCHMARK NEUTRON FIELDS

FOR REACTOR DOSIMETRY

Standard Neutron Field Entries

James A. Grundl

Charles M. Eisenhauer

U S. DEPARTMENT OF COMMERCE

National Bureau of Standards

Center for Radiation Research

Gaithersburg, MD 20899

January 1986

Prepared in cooperation with

U.S. Nuclear Regulatory Commission

Washington, DC 20555

U.S. DEPARTMENT OF COMMERCE, Malcolm Baldrige, Secretary

NATIONAL BUREAU OF STANDARDS, Ernest Ambler, Director

PREFACE

Neutron dosimetry for nuclear technology is a measurement discipline long on

experience and short on major advances. Called upon to establish neutron

exposure parameters in the most diverse kinds of radiation environments and

over a wide range of neutron fl uences, the basic techniques employed have

changed little since early investigations were undertaken to monitor property

changes in materials exposed to neutrons from fission chain reacting systems.

Small passive integral detectors, for the most part activation foils, are the

backbone of measurement. They respond to neutrons over a wide energy range,

often limited by a characteristic reaction threshold. The distinctive

features of energy dependence among the various types of detectors form an

unavoidably complex measurement base for neutron exposure estimates. An

assortment of cal cul ati onal methods, some of trivial sophistication and others

of great ingenuity, has been used to extract relevant neutron exposure

parameters from these specialized integral measurements. Despite the many

limitations in this measurement method, passive integral detectors remain the

technique of choice for neutron dosimetry because of their unmatched flexibility in application and their ability to distinguish neutrons from other

types of radiation.

The inherent ambiguity of integral detector data, as well as the complexity of

the underlying detector response measurements, places a special premium on

measurement experience. Specifically, results from a particular set of

passive neutron detectors often benefit from supporting measurements in one or
more well understood neutron fields, called benchmarks. Neutron fields under

study, which are less well understood and for which neutron exposure

parameters are sought, will always profit from measurements in these benchmarks even when the benchmark and study fields are not well matched.

Nevertheless, similarity of the fields is a benefit and for this purpose a

variety of neutron fields have been developed, or pressed into service, in

order to improve the accuracy of neutron dosimetry measurement techniques.

The present version of the Compendium of Benchmark Neutron Fields for Reactor

Dosimetry is a revision of the original which was distributed in 1978. It is

devoted to three benchmarks called Standard Neutron Fields because they are

simple, well characterized, and remain available at relatively permanent

installations. The Compendium includes a detailed description of each field,

its availability for calibration, calculated detector response parameters

(cross sections and energy response ranges), and a set of observed reaction

cross sections based on an evaluation of experimental data published before

July, 1985. Neutron spectrum and integral cross section errors are included

for each standard field and, at the expense of some unwelcome complexity,

ex pi ai ned

.

For purposes of clarity, space is also devoted to neutron reaction rate formulations, error propagation, and necessary distinctions between measured and

calculated detector response parameters. These formulations are included

because there is no consensus in these matters among neutron dosimetri sts. No

attempt is intended to replace or synthesize the well worn and sometimes

isolated methods by which dosimetry data is interpreted.

TV
TABLE OF CONTENTS

Page

PART IA: 252Cf SPONTANEOUS FISSION NEUTRON SPECTRUM 1

1. IDENTIFICATIONS 3

2. SUMMARY INFORMATION 4

3. FACILITY DESCRIPTION 5 4. NEUTRON FIELD CHARACTERIZATION 8

5. INTEGRAL DETECTOR RESPONSE 12

6. LIST OF TABLES 24

7. FIGURES 39

PART IB: 235U THERMAL-NEUTRON- INDUCED FISSION NEUTRON SPECTRUM .... 41

1. IDENTIFICATIONS 43

2. SUMMARY INFORMATION 44

3. FACILITY DESCRIPTION 45

4. NEUTRON FIELD CHARACTERIZATION 49

5. INTEGRAL DETECTOR RESPONSE 53

6. LIST OF TABLES 58

7. FIGURES 70

PART I

I

A : INTERMEDIATE-ENERGY STANDARD NEUTRON FIELDS ( ISNF)

WITH THICK i0 B SHELL 73

1. IDENTIFICATIONS 75

2. SUMMARY INFORMATION 77

3. FACILITY DESCRIPTION 78

4. NEUTRON FIELD CHARACTERIZATION 81

5. INTEGRAL DETECTOR RESPONSE 96

6. LIST OF TABLES 101

7. FIGURES 118

REFERENCES 125

V



Part IA: 252Cf Spontaneous Fission Neutron Spectrum

1
TABLE OF CONTENTS

Page

1.

IDENTIFICATIONS 3

l.a. Title, Classification

l.b. Designation

l.c. Entry Dates and Revisions

l.d. Generic Description

l.e. Facility Locations

l.f. Contacts for Information

2. SUMMARY INFORMATION 4

3. FACILITY DESCRIPTION 5

3. a. Configuration and Characteristics

(location, source, scattering corrections)

3.b. Irradiation Procedures

(experimental arrangements, neutron fluence)

3.

c. Specification For Transport Calculations

4. NEUTRON FIELD CHARACTERIZATION 8

4.

a. Neutron Fluence

(fluence rate, compensated beam geometry)

4.

b. Neutron Spectrum

4.b.l. Calculation

4.b.2. Measurement

4.

b.3. Evaluated Spectrum

(multigroup format, spectrum errors)

5. INTEGRAL DETECTOR RESPONSE 12

5.

a. Calculated Reaction Probabilities

(formulation, truncation, spectrum fraction)

5.

a.l. Spectrum Response Table

(cross sections, spectrum fraction, response range)

5. a. 2. Spectral Indexes

5. a. 3. Error Propagation

5.b. Measured Reaction Probabilities

(formulation, spectral indexes, neutron fluence transfer)

5.c. Measured Cross Sections and Cal cul ated-to-Observed Ratios

6. LIST OF TABLES 24

7. FIGURES 39

2
1. IDENTIFICATIONS

l.a. NEUTRON FIELD : 252Cf SPONTANEOUS FISSION SPECTRUM

CLASSIFICATION : STANDARD NEUTRON FIELD

l.b. DESIGNATION : XCF-5-N1

NBS 1975 evaluation of documented differential spectrum

measurements [Gr 75b ] and [Gr75c]

l.c. ENTRY DATE : May, 1978 REVISIONS : July, 1985

July, 1983

l.d. GENERIC DESCRIPTION

A standard 252Cf neutron field consists of neutrons from the spontaneous

fission of 252Cf with little or no energy degradation from collided fission or

other background neutrons. The median energy of the spectrum is 1.68 MeV,

with 98% of the neutrons between 0.1 MeV and 8 MeV. Neutron fluence rates in

the range of 10 7 n/(cm 2 s) are obtained in isolated environments near small,

intense 252Cf fission sources. Neutron fl uences are established in terms of

neutron source strength, irradiation time, and source-detector distance; no

microscopic nuclear data or irradiation monitor are required. Certified freefield fl uences of up to 10 13 n/cm 2 may be obtained with uncertainties as low

as ± 1.3% (la).

Measurements of the 252Cf fission neutron spectrum and its close relative the

235 U fission spectrum are extensive and wel 1 -documented . These two standard

neutron fields, therefore, are 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

f i ssi on- spectrum-1 i ke components.

3
l.e. FACILITY LOCATIONS : National Bureau of Standards (NBS)

Center for Radiation Research

Gaithersburg, Maryland 20899

U.S.A.

Contact: Dale McGarry or James Grundl

Phone: 301-921-2767

SEFOR Calibration Facility

University of Arkansas

Fayetteville, AR 72701

U.S.A.

Contact: Leon West

Phone: 501-575-3449

l.f. CONTACT FOR INFORMATION :

Dale McGarry or George Lamaze

Center for Radiation Research

National Bureau of Standards

Gaithersburg, Maryland 20899

U.S.A.

Phone: 301-921-2767

Physi kal i sch-Techni sche

Bundesanstal t (PTB)

33 Braunschweig

Federal Republic of Germany

Contact: Wolfgang G. Alberts

2. SUMMARY INFORMATION

2. a. AVAILABLE FISSION NEUTRON EXPOSURES (NBS ONLY)

• fl uence rate (5 cm from source) ~ 2 x io ; n/(cm 2 s)

o nominal maximum fl uence 1 x 10 13 n/cm 2

® accuracy of free-field fl uence (la) ± 1.3%

2.b. CORRECTIONS FOR NEUTRON SCATTERING AT 5 CM POSITION (NBS ONLY)

9 Net perturbation for threshold detector reaction rates: < (1.0 ± 0.8)%

• Net perturbation for 235U fission reaction rate: (1.2 ± 0.4)%

4
3. FACILITY DESCRIPTION

Fission neutron irradiation facilities employing intense 252Cf spontaneous

fission sources exist at the National Bureau of Standards, the SEFOR Calibration Center in Arkansas, and the PTB Standards Laboratory in the Federal

Republic of Germany. Only the NBS facility which is in general use for

neutron detector calibrations and cross section measurements will be described

in detail. Distinguishing characteri sties of other irradiation facilities may

be ascertained from the literature ([Ma79a], [A175a], [A175b], [Gr75a],

[Br81a] , [Br81b] )

.

3. a. CONFIGURATION AND CHARACTERISTICS [Gr77a], [La82a]

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.

At both locations a light-weight source-detector assembly is available for

irradiation of passive and active neutron detectors. The source-detector

assembly with two active fission chambers mounted on a single axis is shown in

Fig. X- 1 . Alternatively, simultaneous exposure in pairs may be performed

along three separate axes.

The californium source capsule shown in Fig. X-2, is made up of a disk-shaped

2 5

2

Cf deposit in an aluminum pellet encapsulated in a thin-walled stainless

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

5
capsule surface is known to ± 0.5 mm based on constraints of fabrication and

x-ray photographs. Newer 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.

Neutron field parameters and error estimates for a nominal 5 cm source-todetector distance are given in Table X-l. The source strength uncertainty of

± 1.1% dominates the composite error of ± 1.3% for the free-field fission

neutron fluence. The irradiation geometry shown in Fig. X-l is termed compensated-beam geometry and refers to the experimental practice of placing

detectors 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

detectors in this case can be expressed in terms of a mean neutron fluence

which is a function of detector separation with very little dependence on

source position. The source and the detectors may be rotated during the

irradiation to further ensure proper spatial averaging of the neutron fluence

and scattering at the detectors.

Scattering Corrections . Neutron fluence perturbations in location B attributable to room return, scattering in the source capsule and in support

structures are given in Table X-2. Corresponding detector perturbations are

listed for a threshold, and a low-energy response neutron detector. The roomreturn component is estimated from calculation matched to response-versusdi stance measurements in which the source is moved while the detectors are

kept at their normal 5 cm position [Li 84a], For detectors with low-energy

response, a large cadmium basket is placed around the source-detector

assembly to absorb room-return thermal neutrons. The fission neutron return

from this basket is less than 0.1%.

6
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-detector distances of 5 cm and

< 0.5% for distances up to 15 cm.

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

scattering in a detector depends upon the mass, arrangement, and material of

the detector. The NBS has codes available which provide correction factors

for isotropic scattering of neutrons in lightly constructed detectors with

cylindrical symmetry. More extensive Monte-Carlo calculations have been

carried out for special detectors, specifically the NBS double fission

chamber. Multiple scattering in more massive detectors are difficult to

estimate and often require auxiliary experiments. Generally, isolated 252Cf

fission neutron fields are most appropriate for accurate, uncluttered

exposures with light weight detectors.

3.b . IRRADIATION PROCEDURES

Detector pairs of similar sensitivity are hung from the mounting ring on

opposite sides and equidistant from the source as shown in Fig. X-l. Alignment and the exact distance between front faces of detectors are determined

with a computer controlled digital cathetometer fitted with a piezoelectric

sensor. After the separation distance is established, the mounting ring

7
placed in the irradiation facility and the 252Cf source is raised from the

storage hole for a timed exposure. Exposure times in the range of a few

minutes to a maximum of one week are available. Packages of detector pairs

may be irradiated simultaneously along three separate axes. No irradiation

monitor is required since californium is a natural neutron source with an

emission rate governed by its nuclear decay constant (2.2% decrease per

month )

.

A mean free-field fission neutron fl uence for detector pairs (i.e., the

fluence in the absence of all neutron scattering effects) is obtained from the

known 252Cf neutron source strength, the detector separation distance, and the

irradiation time — see Section 4a. This fluence can be specified to a maximum

accuracy of ± 1.3% (la).

3.c. SPECIFICATION FOR NEUTRON TRANSPORT CALCULATION

Neutron transport calculations are employed for scattering corrections.

Material specifications for the source are indicated in Section 3a and in

Fig. X-2.

4. NEUTRON FIELD CHARACTERIZATION

The energy spectrum of 252Cf spontaneous fission neutrons is similar to that

of 235 U 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 at the

same time, 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.

8
upper bound

E p

(p=0.01)

lower bound median

E

p (p=0.99) E

p (p=0.5)

252Cf 0.09 MeV 1.68 MeV

235 U 0.08 1.57

7.8 MeV

7.2

and in a coarse seven-group display, 4>( E ) aE , as follows:

0 0.25 0.8 1.5 2.3 3.7 8 12 MeV

252Cf (0.047) 0.184 0.220 0.194 0.200 0.146 0.009

235 U (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 4.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 X-l, the neutron-to-gamma ratio is about

1.1 x 10 5 (n/cm 2 s)per R/h at 5 cm.

4. a. NEUTRON FLUENCE (see also Section 3. a.)

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

are ~ 6 x 10 9 n/s when fabricated and decays at a rate of 2.2% per month. For

a neutron fluence rate of ~ 2 x lo 7 n/(cm 2 s) at the 5-cm distance, the

fl uence for a one week exposure is ~ 1 x 10 13 n/cm 2

. Spatial gradients for

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

-^ = ^7

- . Free-field fluence accuracy and detector response perturbati ons

for typical irradiation arrangements are discussed in Section 3. a.

9
In the case of compensated-beam geometry where nearly identical detectors are

placed on opposite sides of the source and nearly equidistant from it, a mean

neutron fluence rate may be defined,

S

4ir(r/2) 2

S = source strength (n/s)

r = distance between detectors (cm).

which is proportional to the geometric mean of the detector responses per

atom:

[The exact form of the above expression departs by < 0.2% for D,/^ within 20%

of D

2 /N2 .]

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 perpendicul ar to the capsule axis is 0.992 when

the detectors do not subtend an angle of more than about 20° at the source

(see Fig. X-l)

.

4.b . NEUTRON SPECTRUM

4.b.l. Cal cul ati on . Fission neutron spectra are most accurately determined by laboratory measurement. (See, however. Refs. [Ma82b], [Ma84c] and

earlier work cited therein.)

where D/N = detector response per atom

a = reaction cross section

a = efficiency factor (same for both detectors).

10
4.b .2. Measurement [Gr75b], [Gr75c]. From 1952 to the present, well -

documented measurements of 252Cf and 235U fission neutron spectra have

employed every perfected technique of neutron spectrometry [ Wi 70a ]

.

Accordingly, fission spectra are the most widely studied fast-neutron energy

distribution in existence. The large body of documented fission spectrum data

was evaluated at NBS in 1975 [Gr75c]. 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 evaluation are based on the sixteen selected measurements listed in

Table X-3.

4.b.3. Evaluated Spectrum . The NBS fission spectrum evaluation is

chosen for this compendium. The 252Cf spectrum is defined up to 20 MeV by

means of a reference Maxwell iam, M(E), modified by four piecewise continuous

linear segments below 6 MeV plus one exponential segment above 6 MeV. The

reference Maxwellian is for the 252Cf fission spectrum is

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

with adjustment functions y(E) as follows:

Energy Interval (MeV)

0.0 - 0.25

0.25 - 0.8

0.8 - 1.5

1.5 - 6.0

6.0 -20

1 + 1.20E - 0.237

1 - 0.14E + 0.098

1 + 0.024E - 0.0332

1 - 0.0006E + 0.0037

1.0 exp[-0.03(E-6.0)/l .0]

11
The evaluated spectrum then is given by x(E) = u(E) • M(E). A 45-group

tabulation of the evaluated 252Cf fission spectrum is given in Table X-4 along

with the corresponding 235U evaluated spectrum (see Part I-B, Section 4.b).

For other energy group structures this tabulation may be interpolated as

recommended in the footnote of Table X-4.

Spectrum uncertainties in the evaluation are based on departures of subsets of

measured data from the final adjusted Maxwellian. This estimate, carried out

in a seven group structure, includes both la and 2c errors. Results are given

in Table X-5.

5. INTEGRAL DETECTOR RESPONSE

Neutron dosimetry measurements with integral detectors generally benefit from

the use of benchmark neutron fields to calibrate measurement techniques and

to reference data interpretation methods. In order to carry out such measurement assurance procedures, and in particular for estimating uncertainties, it

is necessary to distinguish between calculated and measured reaction probabilities and to identify parameters which characterize the energy response of

each detector in the various spectra to which it is exposed. A formulation

which meets these requirements including expressions for first-order error

propagation is presented briefly in Sections 5. a. and 5.b. Tables of

calculated response parameters list cross sections, energy response ranges,

and spectral indexes for a number of integral detectors. An evaluation of

measured reaction cross sections, discussed in Section 5.C., ends with a

comparison of measured and calculated values including an error propagation

based on fission spectrum uncertainties given in Table X-5.

12
5. a. CALCULATED REACTION PROBABILITIES

The calculated reaction probability, R , for an integral detector is equal to

the product of the neutron f 1 uence to which it has been exposed and the

spectrum-averaged reaction cross section:

R

c

= a • $ ( X — 1

)

$ = total spectrum integrated neutron fl uence, (n/cm 2 ).

oo a = spectrum-averaged reaction cross section: / a(E)^(E)dE,

o

where ^(E) is the neutron spectrum normalized to unity. (In many

applications the lower limit of integration is 0.4 eV, the cadmium

cut-off.

)

The reaction probability is sometimes referred to as the total reactions per

target nucleus or the time-integrated reaction rate. In this formulation all

experimental quantities such as detection efficiencies, time history

corrections, and neutron field perturbations are relegated to the expression

for the measured reaction probability given in Eq. (X-10) of Section 5.b.

Integral detector response also may be expressed in terms of an average

neutron fl uence rate, < 4>> = $/T, where T is the effective irradiation

interval. In this case, the calculated quantity of interest (experimentally,

the saturated specific activity for activation detectors) is the average

reaction rate: a<4>>.

13
When neutron spectra extend over a large energy range, a truncated neutron

fluence, [ ip( > E

0

)

• $], may be defined where ^(> E

0

) is the spectrum fraction

above E

0

. An example from materials damage dosimetry is [^(> 1 MeV) • $], the

fluence greater than 1 MeV. Similarly, for integral detectors with a

restricted energy response range (e.g., threshold detectors), a cross section

truncated near the lower bound of the detector response range is a measure of

detector response invariance to spectrum shape:

oo

/ a(EH(E)dE

a(> E ) =

1 PJ OO , E = truncation energy . (X-2)

P

For a fractional response p, the truncation energy is defined by

oo P

• 5 = / a(E)i-(E)dE = a(> E.) • (> Ej

r K P

( X-3)

P

where iH> E ) = the spectrum fraction above E

P P

p = fraction of detector response above E ; Ep(p=l) = 0.4 eV; Ep(p=0) = 20 MeV; Ep(p=0.5) = median energy.

The advantage of a truncated cross section cr(> E ) is its independence of

spectrum in the energy region where the detector does not respond.

In terms of these energy-truncated quantities the calculated reaction

probability of Eq. (X-l) may be written as

( X-4)

14
An alternative formulation of truncated fluence which deals with o(> E

0

) in

place of a(> E ) in Eq. (X-4) has been in some use in the past. The relationship between the two formulations follows directly from the definition and

Eq. (X-4),

-, F

x _ freaction probability!

0' " [fluence above E

0 ]

E o>

P

* <K> E

0

) a(> V

5.a.l. Spectrum Response Table . Basic integral detector response

parameters for 252Cf fission neutrons are given in Table X-6 (B5). Spectrum

averaged cross sections (above a cadmium cut-off energy of 0.4 eV), listed in

column 2, are followed in column 3 by cross sections truncated at p

= 0.95.

The energy dependent cross sections employed are those of the ENDF/B-V

dosimetry file reduced to 620 energy groups; spectrum averaging is carried

out with the NBS DETAN code. Spectrum fractions for each detector, [\^(> E ),

p

= 0.95], are listed in column 4. Energy response characteristics given in

the last three columns, are the median response energy and lower- and upperenergy bounds that include 90% of the detector response. The lower-energy

bound, Ep(p = 0.95), is the truncation energy corresponding to the truncated

cross section given in column 3.

5. a. 2. Spectral Indexes . Elementary spectrum indicators associated with

integral detector responses are the ratios of spectrum-averaged cross sections

among detector pairs with distinguishable energy response ranges. This ratio

is generally called a spectral index:

S a/3 a

c/ a

3

’

where a and 3 refer to the two detector reactions involved.

15
A selected set of spectral indexes calculated for ful 1 -spectrum-averaged cross

sections is given in Table X-7(B5). The correspond!' ng set for truncated cross

sections is in Table X-8(B5).

The ratio of the calculated spectral index to the observed value, is the

basic spectrum information available from an integral detector measurement.

These double ratios should be established consistently both for benchmark

neutron fields and for neutron fields under study. Then, by a variety of

analytic means, the double ratios may be employed for (1) directly adjusting

the calculated spectrum of the neutron field under study; (2) normalizing

neutron transport calculations; (3) establishing spectrum consistency among

the benchmark fields themselves; or (4) validating or adjusting reaction

cross sections. In particular, a benchmark neutron field calibration may be

established by normalizing the double ratio for a study spectrum to the

corresponding double ratio for a relevant benchmark:

5. a. 3. Error Propagation . Apart from the complexities of assessing

correl ations, expressions for estimating the uncertainty of calculated

spectrum response parameters can be obtained directly from first-order

statistical propagation. Four error propagation expressions for uncorrelated

errors in spectrum and cross section uncertainties are included here. Others

may be found in Ref. [Gr82a] along with an example of their application to

reactor pressure vessel irradiation surveillance dosimetry.

(X-5)

(X-6)

16
1. A full -spectrum-averaged cross section, a, in multigroup format is

a = o

Q

• l S. i|>. AE. / I AE . ,

i

•

i

where = OqS^ represents the energy dependent reaction cross section and ij>.

the neutron spectrum. The standard error for a, without covariance, is

(6a) 2 = I

i S' * (S'“*<>" * (S' ( 60

q

) 2

The fractional standard error, 6a/ a, follows directly:

a

.

— i|/. AE.

- y

1 1 a

( X-7)

The quantities 6i|>./i|;. and 6s. /s. are the fractional errors in the normalized

neutron spectrum and the cross-section shape, respectively.

2. The fractional error for a spectral index, S

a/e

a /a e is similar,

S s-)

2

+ I 4E

f

) + ( X-8)

3. The fractional error corresponding to a cross section ratio for one

detector in two different neutron fields, S. , = o./a , is

b/s b s

17
%i -

a.

l

r Sl (&0

2

+ 11-

a.

l (i ^i) iM) c*i iE

i

'6iJ; A 2

y S1 \

( X-8a)

r si

4. The truncated spectrum-averaged cross section in multigroup form is,

a ( > E J ‘ °0 1 S

i ^

AE

i / 1 ^

AE

w i i>i / i >i* P P

and the fractional error is

F

6a( > E )"] P & 1 LaO E

p

)

J

-

a

i 6S. _2

1

1 50 E p

) s

i.

+ 1 i>i

1 - a

i a(> E ) P J

t kt * ( X— 9

)

where u . = i|>. aE ./^( > E ) . l r

i l Y p

5.b . MEASURED REACTION PROBABILITIES

The measured reaction probability, R, is proportional to some detector

response mechanism, D:

R

= *[e * y(N,Y,F, ...)] • D , (X-10)

18
D = observed integral detector response during or after irradiation

(e.g., average count rate from a fission chamber, or counts per

second at the end of irradiation from a gamma counting system, or

total number of permanently registered nuclear particle tracks

divided by T, the length of the irradiation)

.

G(X,t) = activation decay rate factor, (s -1

)

where <j>(t) is the irradiation time history of the neutron fluence

rate, T is the length of irradiation, and X is the decay constant.

For an uninterrupted irradiation of length T at constant fluence

rate, G = [1 - exp(- XT)]/T. The dimensionless quantity C = G/X is

sometimes used in place of G. For detectors that register reaction

rate directly (e.g., fission chambers), G becomes 1/T.

N = number of detector atoms.

e = detection efficiency factor

y = composite of response factors exclusive of e/NG required to connect

observed counts per sec or its equivalent to reaction probability;

e.g., effective fraction of detector atoms including neutron self

absorption corrections (N), fission yield (Y), neutron field perturbations (F), isotopic abundance, branching ratio, extrapolation of

pulse height distribution, effects of competing reactions, background

subtraction, pulse losses, etc.

x

19
Benchmark referencing tests the combined validity of many of the factors

included within the bracket of Eq. (X-10). In dosimetry experiments where

absolute reaction probabilities are not determined, or are separately

interpreted, a comprehensive calibration factor may be established on the

basis of a standard neutron field exposure. In this way, most of the response

factors in Eq. (X-10), including the number of detector atoms, are either

eliminated or their error contribution reduced.

When the specific purpose of benchmark referencing is to determine a neutron

fluence in a neutron field under study, the calibration procedure is referred

to as neutron fluence transfer and may be expressed as follows,

$ study _ R study # abmk _

R study # abmk ^ ^ # ^mk ^ ^

^ ^

^bmk ^bmk

astudy ^bmkastudy^

^

^study^

^

by substituting the observed reaction probability, R, from Eq. (X-10) for

in Eq. (X-l). In many cases of interest the reaction probability ratio is

equivalent, or nearly so, to the ratio of the observed integral detector

responses, , with consequent reduction of uncertainties. The

uncertainty in the cross section ratio will depend upon spectrum differences

between the benchmark and study neutron fields as described by Eq. (X-8a).

The important observed quantities for reactor dosimetry are the ratios of

reaction probabilities among detectors exposed, or corrected, to an equivalent

to total fluence. According to Eq. (X-l), these are simply spectrum-average

cross section ratios, and when they are sensitive to spectrum shape, they are

20
called spectral indexes as noted in Section 5. a. 2. An observed spectral index

is obtained from the measured reaction probabilities of two detectors by

replacing the calculated reaction probabilities in Eq. (X-l) with the measured

values from Eq. ( X— 10) (normalized to the same neutron fluence) and dividing:

^ S

a/3^ obs

= (R a/R

3J

obs

( X-l 2

)

5.c. MEASURED CROSS SECTIONS AND CALCULATED-TO-OBSERVED RATIOS

For the two fission spectrum benchmarks described in this compendium, the

primary integral detector quantities of interest are cross sections and

spectral indexes taken with respect to the 238 U(n,f) reaction. For a natural

neutron source such as 252Cf, free-field neutron fluences may be established

directly and absolute integral cross sections may be determined without

recourse to microscopic nuclear data. A substantial amount of such data

exists in spite of the relatively low fluence rates available. Data for the

252Cf fission spectrum are reported and summarized in references [Ad77a],

[A1 75a] , [A1 75b] , [Da78a], [FI 77a] , [Gi75a], [Gi77a], [Gr75a], [Gr82b],

[Gr83a] , [He75a], [He76a], [Ma78a], [Ma81a], [Ma82a], [Ma83a], [Ma84a],

[Sp77a], [Wa79a], and [ Wi 81a ] . Collections of evaluated measured cross

sections may be found in references [Fa75a], [Fa78a], [Ma80a], [Ma81a],

[Ma82a] and [Ma83a].

A new and restricted evaluation of cross-section measurements reported by June

1983, was performed at NBS for detector reactions considered important for

reactor technology. Results are given in Table X-9 and X - 1 1 . The spectral

21
indexes in Table X-ll are based on true ratio measurements or when necessary

on independent measurements of each reaction. The former is most common for

the fission reactions. The uncertainties given in the Tables are one standard

deviation and include estimates of systematic errors.

Cal cul ated-to-observed ratios of reaction cross sections are shown in

Table X-10(B5). The errors in calculated cross sections correspond to 252Cf

fission spectrum uncertainties from Table X-5 propagated according to the

second term of Eq. (X-7) in Section 5. a. Two reactions, 47Ti(n,p) and

232Th(n,f), show cal cul ated-to-observed discrepancies in excess of 10% and six

of the eighteen reactions, some of considerable significance for reactor

dosimetry and neutron fluence monitoring, are discrepant by 5% or more. The

standard deviation of the seventeen C/E ratios in Table X-10(B5), excluding

47Ti(n,p) and 232Th(n,f), are ± 0.025 about an average departure from unity of

1.034.

Cal cul ated-to-observed ratios of spectral indexes, C /llo , are presented in

a/ Uo

Table X-12(B5). The calculated indexes are from Table X-7. Contributing

errors for observed and calculated indexes are listed separately in columns 4

and 5. As is appropriate for cross section validation, propagated errors for

the calculated indexes assess only the contribution of fission spectrum

uncertainties given in Table X-5. The propagation formula for this partial

error is that of the second term of Eq. (X-8), Section 5. a. 3.

22
The C ratios in Table X-12(B5) are a further indication of how calibration

a/Uo

in a benchmark neutron field can improve the accuracy of a neutron fluence

measurement performed with integral detectors. For example, the C value

for 47Ti(n,p) of 1.29 suggests a serious problem with reaction cross section

data, and the value of 0.946 for 235U fission is a significant

disagreement considering the amount of cross section data which exists and the

accuracy which is expected of the two fission reactions involved . Without

benchmark calibration, a neutron fluence obtained with these detectors would

be biased to the extent of the departures from unity.

23
LIST OF TABLES

See

Section

Table X-l 252Cf Fission Neutron Field Parameters and Error

Components 3 .a

Table X-2 Typical Neutron Fluence and Detector Response

Perturbati ons in Location B Due to Neutron Scattering

(5-cm distance) 3 .a

Table X-3 Differential Spectrum Measurements for NBS Evaluation

of 252Cf and 235 U Fission Neutron Spectra 4.b .2

Table X-4 Evaluated Fission Neutron Spectra for 252Cf and 235 U in 45-Group Format 4.b.3

Table X-5 Error Estimates for 252Cf and 235 U Evaluated Fission

Neutron Spectra 4.b. 3

Table X-6(B5) Integral Detector Response Parameters 5.

a

Table X-7(B5) Calculated Spectral Indexes 5. a.

2

Table X-8(B5) Calculated Spectral Indexes — Truncated 5. a.

2

Table X-9 Observed Integral Cross Sections 5.c

Tab! e X-10 Cal cul ated-to-Observed Ratios of Cross Sections 5.c

Table X-ll Observed Spectral Indexes 5.c

Table X-12(B5) Cal cul ated-to-Observed Ratios of Spectral Indexes —

NBS Evaluation 5.c

24
TABLE X-l. 252Cf FISSION NEUTRON FIELD PARAMETERS AND ERROR COMPONENTS

Free-field fission neutron fluence rate

(5-cm distance; source 6 x 10 9 n/s) 2 x 10 7 n/(cm 2

s)

Source decay rate 2.2% / month

Typical maximum fluence 10 13 n/cm 2

Source capsule and support scattering

(inelastic plus net elastic inscatter)

< 1.2%

Gamma-ray exposure at 5 cm (2.8 yr after separation)

Errors in the free-field fission neutron fluence (la):

~ 170 R/h

Source strength ± 1.1%

Source capsule and support scattering ± 0.4% (max .

)

Distance measurements

(typical for compensated-beam geometry)

± 0.5% (max .

)

Quadrature sum: ±1.3% (max.)

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

may not absorb.

TABLE X-2. TYPICAL NEUTRON FLUENCE AND DETECTOR RESPONSE PERTURBATIONS IN LOCATION B DUE TO NEUTRON SCATTERING (5-cm DISTANCE)

Neutron fluence perturbations:

(above 0.4 eV)

room return source capsule scattering

support structure scattering

air scatter

0.1%

0.8%

(0.2 - 0.4)%

< 0.1%

Net perturbation due to neutron scattering:

235U(n,f) detector with cadmium cover:

room return

source capsule

support structures (maximum)

(0.1 ± 0.01)%

(0.8 ± 0.4)%

(0.3 ± 0.1)%

238U(n,f) threshold detector:

room return

source capsule

support structures

(0.0 ± 0.01)%

(0.0 ± 0.5)%

(0.3 ± 0.3)%

Source capsule perturbations depend markedly upon detector threshold.

25
TABLE X-3. DIFFERENTIAL SPECTRUM MEASUREMENTS FOR NBS EVALUATION OF

252Cf AND 235U FISSION NEUTRON SPECTRA

252Cf S P 0 N T A N E 0 US FISSION

Reference

Measurement

Range (MeV) Detection Scheme

[Gr73a] Green et al

.

0.5 - 13 Time-of-fl i ght (TOF); energy scale

based on carbon scattering

resonances. Scintillator NE-213

efficiency from long counter

response (E < 4 MeV); calculations

for E > 4 MeV.

[Gr73a] Knitter et al

.

TOF; energy scale calibrated with

monoenergeti c neutrons (0.4 to

2.3 MeV). Liq. scint. eff. from

angular distribution of T(p,n),

D(d,n), Li(p,n), H(n,n), etc. (0.05

to 8.2 MeV).

[We72a] Werle and Bluhm

(n,p)

( 3 He)

1.5 -

1

74

Proton-recoi 1 proportional counters

(PR); response functions from

monoenergeti c response.

3 He-spectrometer and prop. Gas

counter.

[Gr73a] Meyer et al

.

0.05 - 8 PR (0.05 to 1.2 MeV). Scint. NE-213;

unfolded spectrum (1.1 to 8 MeV).

[Gr73a] Jeki et al

.

0.002- 1 TOF

;

6 Li glass scint. efficiency same

as Meadows.

[We72a] Meadows 0.003- 15 TOF; 6 Li glass scint. (E < 2.6 MeV);

liq. scint. (E > 1.0 MeV)

;

calculated efficiencies.

[We72a] Cond£ and During 0.07 - 7.5 TOF; 6 Li Glass scint. (E < 1.1 MeV),

Calc, eff.; NE-102A plastic scint.

(E > .3 MeV), eff. Measured against

calibrated long counter.

[We72a] Zamyatnin et al

.

0.04 - 6 TOF; 6 Li glass and plastic scint.

26
TABLE X-3. DIFFERENTIAL SPECTRUM MEASUREMENTS FOR NBS EVALUATION OF

252Cf AND 235U FISSION NEUTRON SPECTRA (Continued)

235 U THERMAL-NEUTRON-INDUCED FISSION

Measurement

Reference Range (MeV) Detection Scheme

[Cr56a] Rosen 0.3 - 13 Incident neutron energy (E.j n = thermal. Photographic plate method,

background measurements included.

[ I s 73a] Islam and Knitter 0.55 - 7 E^

n = 0.4 MeV, TOF; energy scale

checked with monoenergeti c neutrons.

NE-102 scint. eff. from H(n,n), up

to 5.75 MeV).

[Cr56a] Barnard et al

.

0.1 - 4 E-jp = 0.1 MeV, TOF; Plastic scint.

calibrated with long ctr.

(E < 2.3 MeV).

[We72a] Cond£ and Duri ng 1 7.5 E-jp = 0.04 MeV, TOF; Plastic scint.

only, same as for 252 Cf by Condd

and During.

[We72a] Werle and

(n.P)

( 3 He)

B1 uhm

1.5 -

1

74

Same as for 252 Cf by Werle and

B1 uhm. E-jp = thermal .

[Cr56a] Cranberg et al

.

0.18 - 2.7 E.j = 0.08 MeV, TOF; Plastic scint.

calibrated against long ctr.

(0.1 - 3.4 MeV).

[Is 73a] Johansson et al

.

0.5 - 14 E-jp = 0.1 - 2.0 MeV, TOF; energy

scale based on scattering resonances

(0.5 - 21 MeV). NE-213 scint. eff.

from H(n,n) and T(p,n) angular

distribution.

[Cr56a] Watt 2.6 - 16 E-jp = thermal. Proton-recoi 1 , gas

counter telescope.

27
TABLE X-4. EVALUATED FISSION NEUTRON SPECTRA FOR 252Cf AND 235 U IN 45-GROUP FORMAT

Designation: XCF-5-N1 Designation: XU5-5-N1

25 2 Cf 235U

Lower

Energy Group Cumulative^ 3 ) Group Cumulative^ 3 ) Boundary FI uence FI uence to Lower FI uence FI uence to Lower

(MeV) 4> • AE Boundary <{> • 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 0.961 5 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 0.871

7

0.0367 0.8582

0.6 0.0343 0.8374 0.0365 0.821

5

0.7 0.0338 0.8031 0.0358 0.7851

0.8 0.0664 0.7693 0.0697 0.7493

1 .0 0.0641 0.7029 0.0668 0.6796

1 .2 0.0608 0.6388 0.0629 0.6128

1.4 0.0290 0.5780 0.0298 0.5499

1 .5 0.0279 0.5490 0.0286 0.5201

1.6 0.0527 0.5211 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 0.041

3

0.1 555 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 0.001 37 0.00471 0.00087 0.00271

9.5 0.00098 0.00334 0.00059 0.00184

10.0 0.00118 0.00236 0.00068 0.00125

11 0.00059 0.00118 0.00031 0.00057

12 0.00030 0.00059 0.00014 0.00026

13 0.00015 0.00029 0.00007 0.00012

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 252 Cf; a = 0.76 for 235 U.

( 3

) Spectrum fraction above lower energy bound.

28
TABLE X-5 . ERROR ESTIMATES FOR 252Cf and 235 U EVALUATED

FISSION NEUTRON SPECTRA

Energy

Boundari es

2 32Cf

(Spontaneous Fission)

Error

Group la

Fluence (%)

<f> • AE

C\J

2 3 5 u (Thermal -NeutronGroup

FI uence

4> • AE

Induced Fission)

Error

la 2 a

(%) (%)

0.0

0.047 ±13 ±26 0.054 ±16 ±32

0.25

0.184 ± 1.1 ± 3.3 0.197 ± 4.1 ± 6.2

0.8

0.220 ± 1.8 ± 3.6 0.229 ± 3.0 ± 4.8

1.5

0.194 ± 1.0 ± 3.1 0.195 ± 3.1 ± 5.2

2.3

0.200 ± 2.0 ± 3.0 0.192 ± 2.0 ± 3.0

3.7

0.146 ± 2.1 ± 4.8 0.127 ± 4.8 ± 8.0

8

0.0087 ± 8.5 ±17 0.0056 ± 5.3 ±11

12

20

( 0 . 00058 ) ( 0 . 00026 )

29
TABLE X-6(B5). INTEGRAL DETECTOR RESPONSE PARAMETERS

Spectrum : NBS-Eval uated 252Cf Spontaneous Fission

Cross Sections: ENDF/B-V Dosimetry File [EN79a]

Designation : XCF-5-N1

Entry Date : May, 1978

Revi sed

:

March, 1982

Detector

Reaction

Cross

o(>

Section^ 3 ) E p)Spectrum^

)

Fracti on *(> E 95)

Median^)

Response

Energy

E p (p=0.5)

Response Ranged

E p =0.4eV E

p

( p=0 . 95

)

E p(p-0.95) E

p (p=0.05)

(barns) (barns) (MeV) (MeV) (MeV)

Spectrum Check

Constant 1 .000 1 .000 0.950 1 .68 0.260 5.5

recip. vel ., (1/v) 0.1107 0.1062 0.991 0.995 0.089 4.3

Fi ssionabl e Mat 1 1 s 239 Pu(n,f

)

1.792 1.807 0.942 1.75 0.286 5.7

235 U( n ,f

)

1 .236 1 .223 0.960 1 .68 0.224 5.8

233 U( n,f

)

1.904 1 .888 0.958 1 .62 0.232 5.5

238 U(n,f

)

0.3136 0.541 0.550 2.79 1 .50 7.2

23 7 Np( n ,f

)

1.352 1 .591 0.807 2.06 0.69 6.2

232Th(n,f) 7.81 E-2 0.1360 0.545 3.01 1.51 7.7

240 Pu(n,f) 1.356 1.598 0.806 2.06 0.69 6.1

241 Pu(n,f) 1.595 1.575 0.962 1.63 0.22 5.6

238 U( n , y) 6.83 E-2 6.55 E-2 0.991 0.90 0.088 2.8

232Th( n, y) 8.97 E-2 8.63 E-2 0.987 0.92 0.109 2.9

Capture Reactions

23 Na 0.2712 E- 3 0.259 E-3 0.996 0.95 0.055 4.7

45 Sc 5.26 E-3 5.01 E-3 0.997 0.65 0.041 2.9

58 Fe 1.660 E-3 1.587 E-3 0.994 1.07 0.067 4.8

59 Co 6.03 E-3 5.76 E-3 0.994 1.08 0.068 4.0

63 Cu 9.65 E-3 9.22 E-3 0.994 0.98 0.067 4.0

115 In 0.1212 0.1169 0.986 1.12 0.117 3.0

19 y Au 0.0763 0.0730 0.994 0.75 0.068 3.0

30
TABLE X-6( B5) . INTEGRAL DETECTOR RESPONSE PARAMETERS (Continued)

Spectrum: NBS-Eval uated 252 Cf Spontaneous Fission

Cross Sections: ENDF/B-V Dosimetry File [EN79a]

Designation : XCF-5-N1

Entry Date : May, 1978

Revised: March, 1982

Detector

Reaction

Cross Section^

»(> Ep )

Spectrum^)

Fraction

'K > ^ 95 )

Median^)

Response

Energy

E

p (p=0.5)

Response Ranged ^

E p =0.4eV E p (p=0.95) E p (p=0.95) E

p (p=0.05)

(barns) (barns) (MeV) (MeV) (MeV)

Helium Production

10 B(n,a) 0.489 0.468 0.992 1 .44 0.078 5.6

6 Li (n , a) 0.465 0.454 0.972 1.55 0.179 6.0

Threshold Reactions

li5 In(n,n‘

)

0.1819 0.270 0.641 2.71 1.20 6.2

4/ Ti (n,p) 0.2407 E-l 0.520 E-l 0.440 3.9 1.92 8.0

32 S(n,p) 0.0760 0.206 0.350 4.1 2.3 7.9

58 Ni (n ,p) 0.1138 0.285 0.380 4.2 2.2 7.9

54 Fe(n ,p) 8.83 E-2 0.249 0.337 4.3 2.4 8.0

46 Ti(n,p) 1.347 E-2 8.85 E-2 0.1445 5.9 3.8 9.9

27 At(n,p) 5.14 E-3 2.79 E-2 0.1749 6.0 3.5 9.8

56 Fe( n ,p) 1.414 E-3 2.79 E-2 0.0482 7.6 5.6 12

63 Cu(n,a) 0.758 E-3 9.72 E-3 0.0741 7.7 4.9 12

27 At(n,a) 1.059 E-3 4.08 E-2 0.0247 8.6 6.6 12

48 Ti (n,p) 0.409 E-3 1.099 E-2 0.0354 8.4 6.0 13

Additions^

i03 Rh( n ,n' ) 0.703 0.849 0.787 2.37 0.75 6.0

93 Nb( n ,n '

)

0.1605 0.220 0.694 2.61 1.02 5.9

60 Ni ( n ,p) 3.44 E-3 4.57 E-2 0.0715 7.3 4.9 11 55 Mn( n ,2n) 0.440 E-3 0.401 1.044 E-3 13 11 16

dpa 895 1030 0.825 2.65 0.64 6.8

31
Footnotes for TABLE X-6(B5)

^The value given in column 2 is the ful 1 -spectrum-averaged cross section

above a cadmium cut-off of 0.4 eV. The truncated cross section on column 3

is for a truncation energy (column 6) above which 95% of the detector

response occurs. A spectrum-averaged cross section truncated at energy E p is given by

a(> E ) = I

o( EH(E)dE / / *(E)dE

(b) The fraction of the spectrum above E 95 = E p (p = 0.95): iK> E 95 ) =

/

i|»(E)dE. The ful 1 -spectrum-averaged cross section g(> 0.4 eV) is equal

^(> E 95 )/0.95.

95

to a(> E 95 >

(c) The fractions p

= 0.95, 0.5, and 0.05 define energies above which 95%, 50%

(median), and 5% of the

defined by the relation

(median), and 5% of the detector response occurs, respectively. E is

P

/ a(E) ip(E)dE = p

• [a(> 0.4 eV)D

where E (p = 1) = 0.4 eV, and E(p = 0) = 20 MeV; and / ^(E)dE = 1.

P 0.4 eV

fdl

18 MeV

' ^Normalization is / a., E^(E)dE = 1.

0.4 eV

i/v

(

6

) Cross sections not taken from the ENDF/BV Dosimetry File:

- Nb(n,n') is from the IRDF Dosimetry File [Cu80a].

- Rh(n,n‘) is from data reported in Ref. [Pa80a].

- dpa, the atom displacement cross section, is from ASTM Standard Practice

E693-79 [An83a]

.

32
TABLE X-7(B5). CALCULATED SPECTRAL INDEXES

Spectrum : 252Cf Spontaneous Fission Desi gnation : XCF-5-N1

(NBS evaluation, [Gr75b], [Gr75c])

Entry Date : Feb. 1980

Spectral Index : Ful 1 -Spectrum Cross Section

from Table X-6(B5) Revised : March, 1982

(ENDF/B-V Dosimetry File)

t S a/3-lcalc. = [^c/^gJcalc.

235U(n,f)

a = 1.236b)

238U(n,f

)

(a = 0.3136b)

N

^

NN\

235U(n,f

)

a = 1.236b)

238U( n ,f

)

(a = 0.3136b)

Fi ssion Threshold

239p u 1.450 5.71 115 In(n,n‘

)

0.1472 0.580

2 3 5 u 1.000 3.941 47Ti (n,p) 0.01947 0.0768

2 3

3|J 1.540 6.07 58 Ni (n,p) 0.0921 0.363

2 3 8y 0.2537 1.000 54 Fe(n,p) 0.0714 0.2816

237 Np 1.094 4.31 46Ti (n,p) 0.01089 0.0430

240p u 1.097 4.32 63 Cu( n, a) 6.13 E-4 2.417 E-3

He Prod. 56 Fe( n ,p) 1.144E-3 4.509 E-3

10 B(n,a) 0.395 1.559 48Ti(n,p) 3.31 E-4 1.304 E-3

6 Li (n, a) 0.376 1.483 27A£(n, a) 8.57 E-4 3.38 E-3

Capture Additions

2 3 8y 0.0553 0.2178 103 Rh(n ,n '

)

0.569 2.242

232Th 0.0725 0.2860 93 Nb( n ,n

1 ) 0.1299 0.511

197Au 0.0619 0.2433

59 Co 4.88 E-3 0.01923

5S Fe 1.343 E-3 5.29 E-3

33
TABLE X-8(B5). CALCULATED SPECTRAL INDEXES - TRUNCATED

Spectrum : 252Cf Spontaneous Fission Designation : XCF-5-N1

Entry Date : Feb. 1980

Spectral Index : Truncated Cross Section

' from Table X-6(B5) Revi sed : March, 1982

(ENDF/B-V Dosimetry File)

[ S a/e]calc. = [°at >E 95^ / ^ 3

^ > E

9 5 ) J cal C .

238 U(n,f)

a( >E g5 ) =0.541b

238 U(n,f)

a(>E 95 )=0.541b

Fi ssion Threshol d

239 Pu 3.34 1 15 In( n,n

1 ) 0.500

235y 2.261 47Ti ( n,p) 0.0961

2 38y 1.000 32 S(n,p) 0.381

237 Np 2.941 58 Ni ( n,p) 0.527

24°p u 2.954 54 Fe( n,p) 0.460

46Ti (n,p) 0.1636

Additions 56 Fe( n,p) 0.0516

103 Rh ( n , n ‘

)

1.569 63 Cu( n, a) 0.01797

9 3 Nb ( n , n

1 ) 0.407 48 Ti ( n,p) 0.0203

27 A£(n, a) 0.0754

34
TABLE X-9. OBSERVED INTEGRAL CROSS SECTIONS

Spectrum : 252Cf Spontaneous Fission Desi gnation : XCF

Entry Date : May 1978

Revised: June, 1983

Reaction

(

2| \ Cross Section

Val ue

( x 1 0 _2/ cm 2)

Medi an Response

Energy,

^ so

( Me V)

Measurement^

References

Threshold Reactions

238 U( n ,f) 326. ± 2.0% 2.8 [Gr83a], [ G i 7 5 a

]

237 Np ( n , f

)

1365 ± 2.0% 2.1 [Gr83a], [Gi 75a]

240 Pu( n,f

)

1337 ± 2.4% 2.1 [Gr83a]

2 3

2

Th ( n ,f) 89.4 ± 3.0% 3.0 [Gr83a]

9 3 Nb ( n , n

1 ) 149 ± 7% 2.5 [A1 82a]

115 In(n,n') 195. ± 1.9% 2.7 [La82a], [Ma79a]

4/ Ti(n,p) 19.4 ± 2.5

%

3.9 [A1 75b] , [Sp77a]

58 Ni ( n,p) 119.4 ± 1.5% 4.2 [La82a], [A1 75b], [FI 77a]

54 Fe(n,p) 87.8 ± 2.2% 4.3 [FI 77a] , [A1 75b] , [Sp77a]

4 6 T i ( n , p

)

14.2 ± 2.5% 5.9 [ A1 75b], [Sp77a]

63 Cu(n, a) 0.696 ± 3.0% 7.7 [Wi81a], [Ma82a]

56 Fe( n,p) 1.45 ±2.5% 7.6 [A1 75b]

4 8 Ti ( n , p

)

0.425 ± 2.5% 8.4 [A1 75b ] , [ Sp 77a

]

27 A£(n,a) 1 .024 ± 2.5% 8.6 [A1 75b], [ Sp 77a

]

Non-Threshol d Reactions

197 Au(n,y) 77.4 ± 2.5% 0.74 [Ma79a] , [Gr75a]

239 Pu(n,f) 1824. ±1.9% 1 .75 [Gr83a] , [Gi75a]

235 U(n,f) 1 21 6. ±1 .6% 1 .68 [Gr83a] , [He76a], [Gr82b],

[Sp80a], [Wa79a], [Da78a]

233 U(n,f) 1893 ± 2.5% 1 .62 [Gr83a]

241 Pu(n,f) 1616 ± 5.0% 1 .63 [Gr83a]

a The total error includes a consideration of the agreement among independent

measurements when they are available, as well as the error estimate given

in the primary reference.

^Reference for the primary measurement in the evaluation is listed first.

35
TABLE X-l 0( B5) . CALCULATED- TO-OBSERVED RATIOS OF CROSS SECTIONS

Spectrum : 252Cf Spontaneous Fission

(NBS Evaluation)

Cross Section: Calculated: ENDF/B-V

Desi gnati on : XCF-5-N1

Entry Date : March 1982

Revised: June, 1983

Reaction

Fi ssion

Cross Section (10~ 27 cm 2

)

Cal cul ated

(

a ) (Table X-6(B5)

Observed

(Table X-9)

239 Pu(n ,f

)

1824 + 1 .9 % 1792 + 0.1 % 0.982 + 1.9

235 U( n ,f

)

1216 + 1 .6 % 1236 + 0.1 % 1 .016 + 1 .6

233 U(n,f) 1893 + 2.5 % 1904 + 0.1 % 1 .006 + 2.5

238 U(n ,f

)

326 + 2.0 % 313.6 + 0.8 % 0.962 + 2.2

237 Np( n ,f

)

1365 + 2.0 % 1352 + 0.6 % 0.990 + 2.1

232 Th(n,f) 89.4 + 3.0 % 78.1 + 0.9 % 0.874 + 3.1

240 Pu( n ,f

)

1337 + 2.4 % 1356 + 0.6 % 1 .014 + 2.5

241 Pu(n ,f

)

1616 + 5.0 % 1 595 + 0.2 % 0.987 + 5.0

Capture

197 Au(n,y) 77.4 + 2.5 % 76.3 + 2.0 % 0.986 + 3.2

Threshold

93 Nb(n,n'

)

149 + 7 % 160.5 + 0.8 % 1.08 + 7%

115 In(n ,n'

)

195 + 1.9 % 182 + 0.8 % 0.93 + 2.1

47 Ti (n,p) 19.4 + 2.5 % 24.1 1.1 % 1 .242 + 2.7

58 Ni (n ,p) 119.4 + 1.5 % 114 + 1.2% 0.955 + 1.9

54 Fe( n ,p) 87.8 + 2.2 % 88.3 + 1 .4 % 1 .006 + 2.6

46 Ti ( n ,p) 14.2 + 2.5 % 13.5 + 2.1 % 0.95 + 3.3

63 Cu(n,a) 0.696 + 3.0 % 0.758 + 3.7 % 1 .09 + 4.8

56 Fe( n ,p) 1 .45 + 2.5 % 1 .414 + 3.2 % 0.975 + 4.1

27 At(n , a) 1 .024 + 2.5 % 1 .059 + 5.3 % 1 .034 + 5.9

48 Ti ( n ,p) 0.425 + 2.5 % 0.409 + 4.7 % 0.962 + 5.3

Cal culated-( b ) to-Observed

Errors correspond to 252 Cf fission spectrum uncertainties (la) given in

Table X-5 and propagated according to the second term of Eq. X-7, Section

5 .a . 3

.

(b) Errors are quadrature sum of errors in columns 2 and 3.

36
TABLE X-ll. OBSERVED SPECTRAL INDEXES

Spectrum : 252 Cf Spontaneous Fission Designation : XCF

Entry Date : May 1 978

t S a/U8lobs

=

/ 5

f (u238) ]obs Revi_sed: June, 1983

. . Reaction^ . . Reactions^ 3 ' Non-overlap Spectral^ 0 ' Interval (MeV) Index

Threshold Reactions (la)

2 3

7

Np ( n,f)/U8 0.7 - 1 .5 4.1 9

+ 1 .5%

240 Pu( n,f)/U8 0.7 - 1 .5 4.10 + 1 .6%

115 In(n,n')/U8 1 .2 - 1 .5 0.598 + 2.8%

47 Ti(n,p)/U8 1 .9 - 1 .5 0.0595 + 3.2%

58 Ni (n ,p) /U8 2.2 - 1 .5 0.366 + 2.1%

54 Fe ( n ,p) /U8 2.4 - 1 .5 0.269 + 3.0%

46 Ti ( n ,p) /U8 3.8 - 1 .5 0.0436 + 3.2%

63 Cu( n,a) /U8 4.9 - 1 .5 2.14 E-3 + 3.6%

56 Fe(n,p)/U8 5.6 - 1 .5 4.45 E-3 + 3.2%

48 Ti(n,p)/U8 6.0 - 1 .5 1 .304E-3 + 3.2%

27 A£(n,a)/U8 6.6 - 1 .5 3.14 E-3 + 3.2%

Non-Threshold Reactions

239 Pu(n,f)/U8 0.3 - 1 .5 5.60 + 1 .2%

235 U(n,f)/U8 0.22 - 1 .5 3.73 + 1 .2%

233 U(n,f)/U8 0.23 - 1 .5 5.81 + 2.1%

241 Pu(n,f)/U8 0.22 - 1 .5 4.96 + 5.7%

197 Au(n,T)/U8 0.06 - 1 .5 0.237 + 3.2%

( 3 ^ All spectral indexes are taken rel ative to the 238 U(n,f) reaction 1 abel ed

"us".

(^Non-overlapping energy response interval between the 95% response energy

for each detector (see Table X-6 and footnote (b) }.

(cl'Fission detector indexes are from simultaneous measurement of fission rates

from both isotopes and hence show smaller uncertainties because of reduced

neutron fl uence errors. Other spectral indexes are formed from independent

cross section measurements listed in Table X-9.

37
TABLE X- 1 2 ( B5 ) . CALCULATED-TO-OBSERVED RATIOS OF SPECTRAL INDEXES - NBS EVALUATION

Spectrum : 252Cf Spontaneous Fission

(NBS evaluation, [Gr75b], [Gr75c])

Spectral Index : Calculated: Table X-7(B5)

Observed: Table X-ll

Desi gnation : XCF-5-N1

Entry Date : March, 1982

Revised: June, 1983

C a/U8 ~ C s a/U8^cal c. 1 [ S a/U8]obS. or C U8/a

Reaction^ a ) Reaction^)

Non-overl ap Interval (MeV)

Cal cul atedTo-Ob served

Val ues ObservedVal ue

ERRORS

Cal cul ated^ 0 ) Val ue

(la)

Total

Threshold Reactions

U8/ 237 Np(n,f) 0.7 - 1.5 0.972 + 1.5% ± 0.5% + 1.7%

U8/ 240 Pu 0.7 - 1.5 0.949 + 1.6% ± 0.5% + 1.7%

47Ti (n,p)/U8 1.5 - 1.9 1.29 + 3.2% ± 0.5% + 3.2%

58 Ni(n,p)/U8 1.5 - 2.2 0.992 + 2.1% ± 0.7% + 2.2%

54 Fe(n,p)/U8 1.5 - 2.4 1.047 ± 3.0% ± 0.8% + 3.1%

46Ti (n ,p)/U8 1.5 - 3.8 0.986 ± 3.2% ± 1.7% + 3.6%

63 Cu(n, a)/U8 1.5 - 4.9 1.129 ± 3.6% ± 3.4% + 5.0%

56 Fe(n,p)/U8 1.5 - 6.0 1.013 ± 3.2% ± 3.0% + 4.4%

48Ti (n,p)/U8 1.5 - 6.0 1.000 ± 3.2% ± 4.1% + 5.2%

2/Ai(n,a)/U8 1.5 - 6.6 1.076 ± 3.2% ± 4.8% + 5.8%

Non-Threshold Reactions

U8/ 239 Pu(n,f) 0.3 - 1.5

5.7 - 7.2

1 ow high

0.981 ± 1.2% ± 0.8% + 1.4%

U8/ 235U(n,f) 0.22- 1.5 5.2 - 6.7

1 ow high

0.946 ± 1.2% ± 1.0% + 1.6%

U8/ 233U(n,f

)

0.23- 1.5

5.5 - 7.2

1 ow

high

0.957 ± 2.1% ± 1.0% + 2.3%

U8/ 241 Pu(n,f) 0.22- 1.5

5.6 - 7.2

1 ow

high

0.975 ± 5.7% ± 1.0% + 5.8%

U8/ 197 Au(n,y) 0.06- 1.5 1 ow

high

0.974 ± 3.2% ± 2.7% + 4.2%

(^Spectral indexes are taken with respect to the 238U(n,f) reaction labeled "U8"

with the lower energy response detector in the denominator. The cal cul ated-toobserved ratio for Of(U238) is 0.962 from Table X-10(B5).

) Non-overl appi ng energy response interval between the 95% response energy boundary

for each detector. Also see footnote (b) Table X-ll. For non-threshold

reactions, two non-overlapping energy intervals are listed: The 95% exclusion

interval is labeled "low" and the 5% exclusion interval "high". See

Table X-16(B5).

(^Errors in the calculated value correspond to 252Cf fission spectrum uncertainties

only, given in Table X-5, and propagated according to the second term of

Eq. (X-8), Section 5. a. 3. The error in 5f(U238) calculated is ± 0.8% (la) from

Table X— 10( B5 ) ; error in the observed value is from Table X-ll.

38
I INDOOR

2.8 m

Figure X-l . 2 52 Cf Fission Spectrum Irradiation Faci’i:.. •

39
SOURCE

GUIDE

TUBE

STAINLESS STEEL

CAPSULE LID ( 0. 59g )

Cf DEPOSIT (~ 3mg

)

2 S04

DISI

1 .4 mm 3

)

(Cf

2 02 S04 DISPLACEMENT

ALUMINUM PELLET

( 6.4 mm dia. x 5.7 mm

LONG ; 0. 48 g )

STAINLESS STEEL CAPSULE

(TYPE 304 •, 7.6mm dia •, 0. 80g )

7. 6 mm

Figure X-2

.

252

Cf Fission Neutron Cource Capsule and Attachments.

40


Part IB: 235U Thermal -Neutron- Induced Fission Neutron Spectrum

41
TABLE OF CONTENTS

Page

1.

IDENTIFICATIONS 43

l.a. Title, Classification

l.b. Designation

l.c. Entry Dates and Revisions

l.d. Generic Description

l.e. Facility Locations

l.f. Contacts for Information

2. SUMMARY INFORMATION 44

3. FACILITY DESCRIPTION 45

3. a. Configuration and Characteri sties

(location, fluence rate, scattering)

3.b. Irradiation Procedures

(experimental arrangements, availability, fluence

monitoring)

3.

c. Specification For Transport Calculations

4. NEUTRON FIELD CHARACTERIZATION 49

4.

a. Neutron Fluence

(intensity, gradients, and scattering perturbations)

4.

b. Neutron Spectrum

4.b.l. Calculation

4.b.2. Measurements

4.

b.3. Evaluated Spectrum

5. INTEGRAL DETECTOR RESPONSE 53

5.

a. Calculated Reaction Probabilities

(formulation, truncation, spectrum fraction)

5.

a.l. Spectrum Response Table

(cross sections, spectrum fraction, response range)

5. a. 2. Spectral Indexes

5.b. Measured Reaction Probabilities

5.c. Measured Cross Sections and Cal cul ated-to-Observed Ratios

6. LIST OF TABLES 58

7. FIGURES ..... 70

42
1. IDENTIFICATIONS

l.a. NEUTRON FIELD : 235U THERMAL-NEUTRON-INDUCED FISSION SPECTRUM

CLASSIFICATION : STANDARD NEUTRON FIELD

l.b. DESIGNATIONS : XU5-5-N1 for NBS spectrum evaluation [Gr75b], [Gr75c]

XU5-9-E5 for ENDF/B-V spectrum evaluation

l.c. ENTRY DATE : June, 1978 REVISIONS : July, 1985

July, 1983

l.d. GENERIC DESCRIPTION

The standard spectrum consists of neutrons from the thermal -neutron-i nduced

fission of 235U. The median energy of the spectrum is 1.57 MeV with 98% of

the spectrum between 0.1 and 7 MeV. Contributions from scattered neutrons and

other backgrounds must be small and well defined if a particular 235U fission

source facility is to qualify as a standard neutron field. Fission neutron

fluence rates of up to ~4 x 10 10 n/(cm 2 s) have been generated in reactor

thermal -col umns with spherical cavities. Source to detector distances are

small in such arrangements and the pure fission neutron field is restricted to

a volume of one to ten cubic centimeters.

Extensive and well documented measurements exist for the 235U fission neutron

spectrum and its close relative the 252Cf fission spectrum. Therefore these

two standard fields are much better known than any other benchmark employed

for reactor dosimetry calibration. In the energy range above 2 MeV, many

neutron fields in and around test and power reactors have f i ssi on-spectrim1 i ke components.

43
l.e. FACILITY LOCATIONS : National Bureau of Standards

Center for Radiation Research

Gaithersburg, Md 20899

U.S.A.

Contact: Dale McGarry or George Lamaze

Phone: 301-921-2767

Research Reactor Institute

Kyoto University

Kyoto, Japan

Contact: Itsuro Kimura

Centre D

1 Etude De L'Energie Nucldaire

SCK/CEN

BOERETANG 200

B-2400 MOL

Bel gi um Contact: Albert Fabry or Reactor Physics Section

Phone: (Oil) 32 14 311801

l.f. CONTACT FOR INFORMATION :

Dale McGarry or James Grundl

Center for Radiation Research

National Bureau of Standards

Gaithersburg, Maryland 20899

U.S.A.

Phone: 301-921-2767

2. SUMMARY INFORMATION

2.

a

AVAILABLE FISSION NEUTRON EXPOSURES ( NBS

• fl uence rate ~ 3.5 x

® nominal maximum fl uence 5 x

• accuracy of free-field fl uence ±

ONLY)

10 10 n/ (cm 2 s

)

10 15 n/cm 2

2.3% (la)

2.b. CORRECTIONS FOR NEUTRON SCATTERING IN SOURCE AND ENCAPSULATIONS (NBS

ONLY)

o neut ron fl uence

o neutron reactions with threshold < 3 MeV

® neutron reactions with threshol d > 3 MeV

< 2.5

%

(0.0 to + 0.6)%

(0.0 to - 1.0)%

44
3. FACILITY DESCRIPTION

Fission neutron sources in spherical cavities are in operation at two reactor

thermal -col umn facilities. At SCK/CEN in Belgium a one meter diameter spherical cavity in graphite provides fission neutron fluences in a cyl i nder-source

arrangement which allows low-energy background components to be established

experimentally [Fa74a], [Fa75b], [Fa82a], [Gi 84a] . The source and detector

access geometry, and its location in a reactor thermal column, are shown in

Fig. X-4. At NBS in the United States, a 30-cm diameter spherical cavity

generates a higher fission neutron fl uence than is available at the SCK/CEN

facility. However, because of the smaller diameter cavity and a disk-shaped

fission source, low-energy backgrounds are more difficult to estimate in the

NBS Facility [Fa72b], [Gr72a], [Mc82a], [Mc84a]. At Kyoto University, a highand low-power fission plate converter system is enclosed in a cubical concrete

shield with ~ 2.5 m wall-to-wall separation [Ki 78a ] . Only the NBS facility

will be described in this compendium.

3. a. CONFIGURATION AND CHARACTERISTICS [Mc82a], [Mc84a], [La82a]

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 2 35U 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 X-3. The source-detector assembly is designed so that the cadmium bo*

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

neutron fl uence rates of ~ 4 x 10 10 n/cm 2 s are obtained at the midplane

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

45
thin disk-shaped detectors with diameters up to 13 mm, axial fl uence gradients

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

midplane. Center-to-edge fl uence ratios are more severe — see Section 4a.

Fission neutron return for spherical cavities in graphite has been studied

extensively by means of neutron transport calculations and by experiment in

connection with the general development of standard neutron fields in

spherical geometry [Ei 77a] , [Fa72b], [Fa75c], [Fa82a], [Gr62a]. 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 fl uence from the fission source disks; for detectors with thresholds

above 1 MeV, the response to the cavity return fl uence 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 SCK/CEN

[Fa82a]. More important is fission neutron scattering in the source-detector

capsule. 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 [Mc82a],

Corrections for activation rates are generally between 0.5% and 3%.

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

because of uncertainties in the graphite return background. The graphite

return correction for the 235U(n,f) reaction, for example, is nearly 10% and

self-shielding of the near-l/E cavity return fl uence by the fission disks and

by a 235U activation foil makes this correction uncertain to a factor of about

one third. The one meter cavity arrangement at SCK/CEN is better suited for

low-energy response detectors [Fa82a].

46
3.b. 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 115 In(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.

The expression which governs this fluence transfer is

$

cvy

^cvv °cf

. ^

*cf

R

cf °cvy

cf

"cvy

°r.f< >Eo>

acvy^>E p)

^ A>E ) xf v p

'

(>E ) x vy

v p

cvy

} cf

cf

where $ = fission neutron fluence and "cvy" denotes the 235U cavity fission

source and "cf" the 252Cf fission neutron calibration fl uence.

Ep = truncation energy (p = 0.95 commonly).

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

(G), truncated reaction cross sections a(> E ), spectrum fractions i|»(>E ), and

the ratio of the monitor foil activity in the cavity to that at the 252Cf

irradiation facility — see Section 5. a in Part IA. Details of the fluence

transfer procedure are given in Refs. [Mc84a], [La84a], 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

47
effectively substituted for 58Ni(n,p) and 115 In(n,n') as the fl uence transfer

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

fission neutron fluence to an accuracy of ± 2.3% (la).

Fission neutron fluences of ~ 3 x 10 15 n/cm 2 may be obtained at the NBS

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

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

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

fission sources.

3.c. SPECIFICATIONS FOR TRANSPORT CALCULATIONS

Neutron transport calculations are required for two types of corrections:

(1) neutron scattering in the fission source disks and in the detector

capsule; and (2) neutron return from the graphite.

The graphite thermal column at the NBS Reactor is effectively infinite for

fission neutrons originating in the cavity. Graphite density is 1.74 g/cm 3

, and the cavity is a well formed sphere 29.8 cm in diameter. The two fission

source disks may be considered as a point source at the cavity center.

Scattering in the source-detector capsule has been established on the basis of

Monte Carlo calculations [Mc82a]. The main scattering components in these

calculations (see Fig. X-3) are the cadmium box (22.5-mm dia., 8.3-mm high,

and 0.076-mm thick), the source disks (16-mm dia. x 0.13-mm thick), and the

foil samples (typically less than 2 g). Although the nickel and indium

fluence monitor foils are in the same geometry as the foil samples, they do

not exactly monitor energy-angle effects for all types of threshold detectors.

The latter effect, along with energy transfer by inelastic scattering, are the

largest uncertainties in detector response corrections for neutron scattering.

48
4. NEUTRON FIELD CHARACTERIZATION

The energy spectra of fission neutrons are very similar for all fissionable

materials and both test and power reactors exhibit fission-spectrum-like

components in the MeV energy range. Using the same description as in Part IA,

the 235 U fission neutron spectrum may be described in terms of a broad energy

range,

lower bound median

E p (p=0.99) E

p (p=0.5)

235U 0.08 MeV 1.57 MeV

and a coarse seven-group flux display, <j)(E)AE, as follows:

0 0.25 0.8 1.5 2.3 3.7 8 12 MeV

235U (0.054) 0.197 0.229 0.195 0.192 0.127 0.006

A detailed description of the spectrum is given in Section 4.b.

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

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

4. a. NEUTRON FLUENCE (see also Sections 3. a. and 3.b.)

The combined source strength for the two fission disks in the thermal column

cavity is approximately 3 x 10 11 n/s. Since this source strength is difficult

to determine accurately, the 235U fission neutron fl uence is established by

upper bound

Ep(p=0.01)

7.2 MeV

49
means of fl uence transfer from the 252Cf fission neutron irradiation facility

to an accuracy of ± 2.3%, exclusive of scattering in detectors — see

Section 3.b. The neutron fl uence at the midplane of the source-detector

capsule for a source disk separation of 12 mm is 0.12 n/cm 2 per source neutron

corresponding to a fl uence rate of ~ 3.5 x io 10 n/cm 2 s with the reactor at

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

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

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

midplane. Radial variations are more severe: center-to-edge fl uence 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 115 In

fl uence monitor included in each irradiation package yields a fl uence averaged

over the radial variation.

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

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

free-field activity attributable to both neutron scattering and cavity return:

total 4 measured _ _acti vi ty_ _ acti vi ty

[i MS<J

This scattering correction depends markedly upon reaction threshold because of

energy-angle correlation in the nearby materials of the encapsulation.

50
Components of , obtained from detailed Monte-Carlo calculations of the

source detector capsule with a representative foil stack, are as follows:

neutron

fl uence changes in threshold detector response

E 95 = 1.2 MeV E 95 = 2.1 MeV E g5 = 6.5 MeV

cavity return +1.2% + 0.17% < 0.1% < 0.1%

encapsul ations +2.5% +(0.5 ± 0.3)% -(0.2 ± 0.3)% -(1.3 ± 0.3)%

(incl . source

disks and cd box)

foil stack +1 . 5% +(0.0 to 0.5)% -(0.4 to 1.0)% -(0.8 to 1.5)%

y sc (%) : +(0.7 to 1.2)% -(0.6 to 1.2)% -(2 to 3)%

The spread of the foil -stack component of y scis 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 error of ± 0.6 is assigned to y expressed in

percent.

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

nearly pure fission neutron fl uences in small cavities. The cavity return

fl uence at the cavity center is about 1.2% of the uncollided fission neutron

fl uence. This return fl uence 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.

51
4.b. NEUTRON SPECTRUM

4.b.l. Cal cul at 1 on . Fission neutron spectra are most accurately

determined by laboratory measurement. (See, however. Ref. [Ma82b] and earlier

work cited therein.)

4.b.2. Measurements . See Section 4.b.2. in Part 1A.

4.b.3. Evaluated Spectrum . The NBS fission spectrum evaluation is

chosen for this compendium because of the consistency of its procedures for

both 235 U and 252Cf fission spectra, and because it includes an articulated

error estimate [Gr75c], [Gr75b]. Cross section comparisons with the ENDF/B-V

235 U fission spectrum shape will be given as a supplement.

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 /E exp(-l .5E/1 .97) , E in MeV,

and the adjustment functions y(E) are as follows:

Energy Interval (MeV)

0.8 - 1.5

1.5 - 6.0

0.25 - 0.8

0 - 0.25 1 + 0.8E - 0.153

1 - 0.14E + 0.082

1 + 0.040E - 0.062

1 + 0.01E - 0.017

6.0 - 20 1.043 exp [ — 0 • 06 ( E - 6.0)/1.043]

52
The evaluated spectrum is given by x(E) = u(E) • M(E). A 45-group tabulation

of the evaluated 235U fission spectrum is given in Table X-4, Part IA. Group

fl uences 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 X-5 of Part IA, are based on departures

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

evaluation. This estimate, carried out in a seven-group structure, includes

both la and 2 a errors.

Of the many descriptions employed for the 235U fission spectrum, the most

common is the Watt-function fit chosen for ENDF/B-V in June, 1979:

iKE) = C exp(-E/a) sinh(/bE)

a = 0.988 MeV b = 2.249 MeV" 1 C = 0.4395

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

Contemporary time-of-fl i ght data was emphasized in the ENDF/B-V evaluation as

compared with the NBS evaluation which considered all wel 1 -documented data.

Differences between the two evaluated spectra, in terms of calculated spectral

indexes, are shown in Table X-14a(B5) and X-14a.l.

5. INTEGRAL DETECTOR RESPONSE

Neutron dosimetry measurements with integral detectors generally benefit from

the use of benchmark neutron fields for calibrating measurement techniques and

for referencing data interpretation methods. In order to carry out these

53
measurement assurance procedures, and in particular for estimating uncertainties, it is necessary to distinguish between calculated and measured reaction

probabilities and to identify parameters which characterize the energy

response of each detector in the various spectra to which it is exposed. A

formulation which meets these requirements is described in Section 5 of

Part IA. Response parameters for a number of integral detectors, including

cross sections, spectral indexes, and energy response ranges, are presented in

Section 5. a. Measured reaction probabilities are given in Section 5.b

followed by comparisons of measured and calculated cross sections in Section

5.c.

5. a. CALCULATED REACTION PROBABILITIES

(See Section 5. a in Part IA for reaction probability formulations, parameters,

and error propagation.)

5.a.l. Spectrum Response Table . Basic integral detector response

parameters for 235U fission neutrons are given in Table X-l 3 ( B5 ) . The

spectrum-averaged cross sections listed in column 2 are the ful 1 -spectrumaveraged values above a cadmium cut-off energy of 0.4 eV and are followed in

column 3 by cross sections truncated at p = 0.95. The energy dependent cross

sections employed are those of the ENDF/B-V dosimetry file reduced to 620

energy groups; spectrum averaging is carried out with the NBS DETAN code. The

spectrum fraction, [ip(> E ), p = 0.95], follows in column 4. Energy response

P characteri sties , given in the last three columns, are the median response

energy and the energy response range, i.e., the lower- and upper-energy bounds

that include 90% of the detector response. The low-energy bound,

E (p = 0.95), is the truncation energy correspondi ng to the truncated cross

P section given in column 3.

54
5. a. 2. Spectral Indexes (See Section 5. a. 2, Part IA for formulations) . A

selected set of spectral indexes calculated for full -spectrum-averaged cross

sections from Table X- 13 ( B5 ) is given in Table X-14(B5); a corresponding set

for truncated cross sections is in Table X-15. All of these spectral indexes

are for the NBS evaluated fission spectrum shape described in Section 4.b.3.

Tables 14a(B5) and 14. a. 1 list spectral indexes for the ENDF/B-V fission

spectrum and compare the two evaluations.

5.b. MEASURED REACTION PROBABILITIES (See Section 5.b, Part IA)

5.c. MEASURED CROSS SECTIONS AND CALCULATED TO OBSERVED RATIOS

For this compendium, the primary integral detector quantities of interest are

cross sections and spectral indexes taken with respect to the 238 U(n,f)

reaction. Available integral detector data for the 235U fission spectrum are

to be found in References [Ac81a], [A182a], [Fa74a], [Fa75a], [Fa75b],

[FI 7 7a] , [Gi84a], [Gr72a], [Gr75b], [Gr83a], [Gr85a], [Ki78a], [La82a]

,

[Le57a], [Ma84b], and [0182a]. Collections of evaluated measured cross

sections may be found in References [Fa72a], [Fa78a], [Gr70a], and [Ha78a] .

A new and restricted evaluation of measurements reported before July 1985, was

performed at NBS for detector reactions considered important for reactor

technology. The evaluation begins with the absolute 235 U(n,f) and 239 Pu(n,f)

cross sections measured for 252 Cf fission neutrons. Since these cross

sections for 252 Cf and 235U fission neutrons are the same to within a conservative upper bound of ± 0.6%, cross section measurements performed relative to

either one in the 235U fission spectrum can be scaled to the 252 Cf cross

section value with very little additional error. For most 235 U fission

spectrum measurements performed at NBS and SCK/CEN, this fluence transfer

technique establishes a fission neutron fluence based upon a 252 Cf neutror

55
source strength — see Section 3a, Part IA. For other measurements not so

normalized, the 115 In(n,n') or 58Ni(n,p) reactions are construed in the

evaluation as the fluence monitors. Results are given in Table X- 16 and X- 18

The uncertainties given at one standard deviation include estimates of

systematic errors.

Cal cul ated-to-observed ratios of reaction cross sections are listed in Table

X- 1 7 ( B5 ) and include comparison with both the NBS evaluated and the ENDF/B-V

fission spectrum shapes. Errors in the calculated cross section correspond t

uncertainties in the fission spectrum (Table X-5) propagated according to the

second term of Eq. (X-7)(Part IA, Section 5. a. 3). Only the 47Ti(n,p),

48Ti(n,p), and 232Th(n,f) reactions show cal cul ated-to-observed disagreements

in excess of 10%; however, twelve C/E ratios are outside of one standard

deviation for the NBS evaluated spectrum. Two threshold reactions important

for reactor dosimetry, 115 In(n,n‘) and 58Ni(n,p), disagree with predicted

values by nearly 10%. Of particular importance for cross section evaluation

is the apparent (8-10)% under prediction of the measured 10B(n,He) reaction

cross section by ENDF/B-V. The standard deviation of fifteen C/E ratios in

Table X-17(B5) for the NBS spectrum, excluding 47Ti(n,p), 232Th(n,f), and

239 Pu(n,f), are 0.038 about an average departure from unity of 1.062. The

standard deviation for the same fifteen C/E ratios in the ENDF/B-V fission

spectrum are 0.029 about an average departure from unity of 1.035.

Cal cul ated-to-observed ratios of spectral indexes, C^g, are presented in

Table X-19(B5). The calculated indexes correspond to ful 1 -spectrum averaged

cross section ratios for the NBS spectrum evaluation from Table X- 14 ( B5 )

.

Contributing errors for observed and calculated indexes are listed separately

56
Errors for the calculated indexes correspond to fission spectrum uncertainties

listed in Table X-5 propagated according to Eq. (X-8) (Part IA,

Section 5. a. 3)

.

The C^

u8

ratios in Tabl e-X19(B5) validate integral detector measurement techniques and detector cross sections. The difficulties with the Ti isotopes,

which was so evident with the ENDF/B-IV dosimetry file, are improved except

for 47Ti where a 27% discrepancy remains. The 9% discrepancy for C( 238U/ 235U)

is consistent with that obtained for 252Cf fission spectrum neutrons (see

Table X-12). Of interest for reactor physics is the disagreement of

(4.7 ± 1.7)% with ENDF/B-V of the spectrum independent 239Pu/ 235U fission

cross section ratio. The corresponding disagreement obtained in the 252Cf

spectrum is (3.7 ± 1.5)%.

57
LIST OF TABLES

See

Secti on

Table X - 1 3 ( B5) Integral Detector Response Parameters 5 .a .1

Table X-14(B5) Calculated Spectral Indexes — NBS Evaluated

Fission Spectrum 5. a.

2

Table X-14a(B5) Calculated Spectral Indexes — ENDF/B-V

Evaluated Fission Spectrum 5. a.

2

Table X-14.a .1 Calculated Spectral Index Ratios: NBS Evaluation

vs. ENDF/B-V Evaluation Fission Spectrum 5. a.

2

Table X- 1 5( B5) Calculated Spectral Indexes — Truncated 5. a.

2

Table X-16 Observed Integral Cross Sections 5.c

Table X- 17 ( B5

)

Cal cul ated-to-Observed Ratios of Cross Sections 5.c

Table X-18 Observed Spectral Indexes 5.c

Table X-19(B5) Cal cul ated-to-Observed Ratios of Spectral

Indexes — NBS Evaluation 5.c

58
TABLE X-13(B5). INTEGRAL DETECTOR RESPONSE PARAMETERS

Spectrum : 235U Thermal Neutron-Induced Fission

(NBS evaluation, [Gr75b], [Gr75c ]

)

Cross Sections: ENDF/B-V Dosimetry File [EN79a]

Desi gnati on : XU5-5-N1

Entry Date : June, 1978

Revised: March, 1982

Detector

Reaction

Cross

a{>

Section^ 3 )

E p)

Spectrum^ ^ Fraction

Median^)

Response

Energy Response Ranged ^

E p =0.4eV E p

( p=0 . 95

)

4*( ^ E

g 5) E

p (p=0.5) E p (p=0.95) E

p

( p

= 0 . 05 )

Spectrum Check

(barns) ( barns) (MeV) (MeV) (MeV)

Constant

reci p. vel . , (1 /v)

d ^ 1.000 1.000 0.950 1.57 0.24 5.1

0.1151 0.1103 0.991 0.921 0.076 4.0

Fi ssionabl e Mat '1

s

239 Pu(n ,f

)

1.786 1.800 0.9424 1.65 0.26 5.2

235U( n,f

)

1.236 1.221 0.962 1.57 0.20 5.2

233U(n,f) 1.912 1.895 0.959 1.51 0.21 5.0

238 U( n,f

)

0.2947 0.532 0.527 2.67 1.48 6.7

237 Np(n,f

)

1.322 1.578 0.7958 1.96 0.670 5.7

2 3 2T h ( n , f

)

0.0724 0.1322 0.520 2.86 1.50 7.3

240 Pu(n,f

)

1.326 1.585 0.795 1.97 0.67 5.6

24

1

Pu( n ,f

)

1.600 1.579 0.963 1.53 0.19 5.1

238 U(n, y) 0.0721 0.0691 0.991 0.87 0.075 2.7

232 Th(n, y) 0.0942 0.0906 0.988 0.88 0.094 2.8

Capture Reactions

23 Na 2.817 E-4 2.69 E-4 0.995 0.845 0.053 4.3

45 Sc 5.64 E-3 5.37 E-3 0.997 0.61 0.035 2.7

58 Fe 1.712 E-3 1.634 E-3 0.995 0.948 0.049 4.5

59Co 6.28 E-3 6.00 E-3 0.995 1.024 0.054 3.8

63Cu 1.008 E-2 9.62 E-3 0.995 0.917 0.055 3.8

1 1 5

i n 0.1266 0.1220 0.986 1.080 0.102 2.9

197Au 0.0810 0.0774 0.994 0.695 0.058 2.8

59
TABLE X-13(B5). INTEGRAL DETECTOR RESPONSE PARAMETERS (Continued)

Spectrum : 235U Thermal Neutron-Induced Fission

( NBS evaluation, [Gr75b], [Gr75c])

Cross Sections : ENDF/B-V Dosimetry File [EN79a]

Designation : XU5-5-N1

Entry Date : June, 1978

Revised: March, 1982

Detector

Reaction

Cross Section^ 3 ) “(> E

p

) Spectrum^)

Fraction

e 95 )

Median^ 0 ) Response

Energy

E p (p=0.5)

Response Ranged

E

p =0.4eV E

p (p=0.95) E p (p=0 .95) E p (p=0.05)

(barns) (barns) (MeV) (MeV) (MeV)

Helium Production

10 B(n,a) 0.499 0.478 0.993 1 .16 0.066 5.2

6 Li (n,a) 0.465 0.455 0.971 1.25 0.167 5.7

Threshold Reactions

ll5 In(n,n'

)

0.1734 0.264 0.624 2.6 1.16 5.9

47 Ti(n,p) 0.02159 0.0487 0.421 3.8 1.87 7.5

32 S(n,p) 0.0676 0.198 0.324 4.0 2.3 7.4

58 Ni (n ,p) 0.1009 0.2645 0.362 4.1 2.1 7.5

54 Fe(n ,p) 0.0778 0.2334 0.317 4.2 2.3 7.6

46 Ti (n,p) 0.01081 0.0806 0.1274 5.7 3.8 9.3

2y A&(n,p) 4.121 E-3 2.522 E-2 0.1553 5.8 3.5 9.3

56 Fe(n ,p) 1.006 E-3 2.446 E-2 0.0391 7.3 5.5 11.2

63 Cu(n , a) 5.40 E-4 7.80 E-3 6.58 E-2 7.3 4.7 11.1

27 A£(n , a) 6.93 E-4 3.54 E-2 1.861 E-2 8.4 6.5 11.9

4 8 Ti ( n , p

)

2.726 E-4 9.21 E-3 2.813 E-2 8.1 5.9 12.2

Additions^

1 03 Rh( n ,n '

)

0.682 0.835 0.776 2.3 0.72 5.7

93 Nb(n ,n '

)

0.1543 0.2159 0.679 2.5 0.99 5.6

60 Ni (n ,p) 2.528 E-3 3.99 E-2 0.0603 7.0 4.9 10.4

55 Mn(n,2n) 0.2016 E-3 0.366 0.523 E-3 12.7 11.1 16

dpa 854 973 0.83 2.5 0.57 6.3

60
Footnotes for TABLE X-13 ( B5

)

( a Hhe value given in column 2 is the ful 1 -spectrum-averaged cross section

above a cadmium cut-off of 0.4 eV. The truncated cross section in column 3

is for a truncation energy (col umn 6) above which 95% of the detector

response occurs. A spectrum-averaged cross section truncated at energy E

p is given by

o(> EJ = / a(E)*(E)dE / / *(E)dE . /

(b) The fraction of the spectrum above E 95 = E

p (p = 0.95): i|>(> E g5 ) =

oo

/ ^(E)dE. The ful 1 -spectrum-averaged cross section a(> 0.4 eV) is equal

^ 9 5 to a{> E 95 ) • ij>( > E qc; )/0.95. 95 '

(c) The fractions p

= 0.95, 0.5, and 0.05 define energies above which 95%, 50%

(median), and 5% of the

defined by the relation

(median), and 5% of the detector response occurs, respectively. E

p is

/ a(E)i|>(E)dE = P

• Ca(> 0.4 eV)]

E

where E (p = 1) = 0.4 eV, and E(p = 0) = 20 MeV; and / ^(E)dE = 1.

P 0.4eV

18 MeV

(d) Normal ization is / a^

v

E ip( E ) dE = 1.

. , 0.4 eV

( e

)

y Cross sections not taken from the ENDF/BV Dosimetry File:

- Nb(n,n‘) is from the IRDF Dosimetry File [Cu80a]

- Rh(n,n‘) is from data reported in Ref. [Pa80a],

- dpa, the atom displacement cross section, is from ASTM Standard

Practice E693-79 [An83a],

61
TABLE X-14(B5). CALCULATED SPECTRAL INDEXES— NBS EVALUATED FISSION SPECTRUM

Spectrum : 235U Thermal Neutron-Induced Fission

(NBS evaluation, [Gr75b], [Gr75c])

Spectral Index : Ful 1 -Spectrum Cross Section

~~ from Table X- 1 3 ( B5

)

(ENDF/B-V Dosimetry File)

Designation : XU5-5-N1

Entry Date : June, 1978

Revised: March, 1982

[Sa/B^calc. - I^o/^gJcalc.

62
TABLE X-14a( B5) . CALCULATED SPECTRAL INDEXES—ENDF/B-V EVALUATED FISSION

SPECTRUM

Spectrum : 235U Thermal Neutron-Induced Fission

(ENDF/B-V Nuclear Data File)

Spectral Index : Ful 1 -Spectrum Cross Section

(ENDF/B-V Dosimetry File)

^a/glcalc. =

/ Alcaic.

o\b

235U( n,f

)

{o = 1.236b)

238 U(n,f)

(a = 0.3052b)

235U(n,f)

3 = 1.236b)

238U(n,f

)

{o = 0.3052b)

Fi ssion Threshold

239p u 1.449 5.87 115 In(n,n '

)

0.1451 0.587

235y 1.000 4.05 47Ti (n,p) 0.01817 0.0736

233y 1.543 6.25 58 Ni (n,p) 0.0850 0.344

238y 0.2469 1.000 54 Fe(n,p) 0.0656 0.2655

237 Np 1.090 4.41 46Ti (n,p) 9.04 E-3 0.0366

He Prod • 56Fe(n,p) 8.38 E-4 3.39 E-3

10 B(n, a) 0.397 1.607 63 Cu(n, a) 4.51 E-4 1.829 E-3

6 Li (n. a) 0.368 1.489 27A£(n, a) 5.82 E-4 2.357 E-3

Capture 48Ti (n,p) 2.280 E-4 9.23 E-4

238 U 0.0568 0.2302 Additions

197Au 0.0633 0.2565 103 Rh(n,n 1 ) 0.564 2.284

59 Co 5.00 E-3 0.02023 93Nb(n,n '

)

0.1292 0.523

58 Fe 1.367 E-3 5.53 E-3 3 2S ( n , p

)

0.0570 0.231

Designation : XU5-9-E5

Entry Date : March, 1982

Revi sed

:

63
TABLE X-14a.l . CALCULATED SPECTRAL INDEX RATIO: NBS EVALUATION VS.

ENDF/B-V EVALUATED FISSION SPECTRUM

Spectrum: 235U Thermal Neutron-Induced Fission

(NBS Eval ./ENDF/B-V)

Spectral Index : Ful 1 -Spectrum Cross Section

(ENDF/B-V Dosimetry File)

[^a/S-lxNBS ^ ^a/g-lxgv

Designation :

Entry Date : March, 1982

Revi sed

:

a\8 235 U( n,f) 238 U(n,f) 235 U(n,f) 238 U( n ,f )

Fi ssion Threshold

239pu 0.997 1.032 115 In(n ,n'

)

0.9671 1.001

235u 1.000 1.035 47Ti (n,p) 0.962 0.996

23 3(j 1.003 1.038 58 Ni ( n ,p) 0.960 0.994

238u 0.966 1.000 54 Fe( n ,p) 0.960 0.994

23/Np 0.982 1.018 46 Ti(n,p) 0.971 1.003

He Prod. 56 Fe( n ,p) 0.971 1.006

10 B( n , a) 1.018 1.053 63 Cu(n,a) 0.968 1.002

6 Li ( n , a) 1.022 1.060 27 Ai(n, a) 0.965 0.997

Capture 48Ti (n,p) 0.970 1.002

238u 1.026 1.064 Additions

1 97 Au 1.036 1.072 103 Rh( n ,n

1 ) 0.979 1.013

3 9 Co 1.018 1.053 93 Nb( n ,n'

)

0.966 1.000

58 Fe 1.013 1.051 32S(n,p) 0.960 0.991

64
TABLE X- 1 5 ( B5 ) . CALCULATED SPECTRAL INDEXES— TRUNCATED

Spectrum : 235 U Thermal -Neutron-Induced Fission

( NBS evaluation, [Gr75b], [Gr75c])

Spectral Index : Truncated Cross Sections

from Table X-13(B5), Column 3

(ENDF/B-V Dosimetry File)

Designation : XU5-5-N1

Entry Date : June, 1978

Revised: March, 1982

[^a/B^calc. - E 95^calc.

a\B

Nv a(>

238 U(n ,f

)

E 95 ) = 0.532b

238 U(n ,f

)

a(> E g5 ) = 0.532b

Fi ssi on Threshold

239p u 3.38 115 In(n ,n

1 ) 0.496

235u 2.295 47 Ti (n,p) 0.0915

2 3 3 u 3.56 32 S(n,p) 0.372

238u 1.000 58 Ni (n ,p) 0.497

23/Np 2.97 54 Fe(n,p) 0.439

240pu 2.98 46Ti (n,p) 0.1515

56 Fe( n ,p) 0.0461

Additi ons 63 Cu(n , a) 0.01466

103 Rh(n,n'

)

1.570 48Ti (n,p) 0.01731

93 Nb(n,n '

)

0.406 27 A£(n,a) 0.0665

65
TABLE X- 16 . OBSERVED INTEGRAL CROSS SECTIONS

Spectrum : 235U Thermal -Neutron-Induced Fission Designation : XU5

Entry Date: March, 1982

Revised: July, 1985

React i on Median ResponseEnergy (MeV)

Cross

Secti on

( xio 27 cm 2

) Measurement

Reference

Threshold Reactions

238 U(n,f)

t

2.7 312 ± 2.3% [Gi 84a] , [Le57a]

237 Np(n,f)

t 1.96 1359 ± 2.1% [Gi 84a]

240 Pu(n,f)"^ 1.97 1332 ± 2.1% [Gi 84a]

232 Th( n ,f ) + 2.9 82.9 ± 3.1% [Gi 84a]

93 Nb( n ,n

1 ) 2.4 164 ± 9 % [A182a], [Ko80a]

115 In(n,n'

)

2.6 190 ± 2.1% [Gi 84a] , [ Fa75b]

,

[Ki 78a]

47 Ti (n,p) 3.8 18.0 ± 3.5% [Ma84b], [Ki78a]

58 Ni (n ,p) 4.1 111 ± 2.4% [Gi 84a] , [La82a]

,

[Fa75a] , [Fa74a]

54 Fe( n ,p

1 ) 4.2 81.7 ± 2.7% [Ma84b], [La82a],

[FI 77a]

46 Ti (n,p) 5.7 11.8 ± 3.5% [Ma84b]

63 Cu(n ,a) 7.3 0.60 ± 7 % [Fa75a] , [Ki78a]

48 Ti(N,p) 8.1 0.307 ± 3.5% [Ma84b], [Ki78a]

2/ A£(n,a) 8.4 0.720 ± 3.5% [Ma84b] , [Fa75a]

Non -Threshold Reactions

239 Pu( n ,f ) + 1.65 [1818 ± 1.9%] [Gi84a], [Mc84a]

235 U(n,f)

+ 1.57 1200 ± 1.9% [Gi 84a]

233 U( n,f)

f

1.51 1949 ± 3.1% [Gi84a]

10 B(n,He) 1.16 540 ± 4.0% [Gr85a] , [0182a]

6 Li ( n ,He) 1.25 455 ± 4.0% [Gr85a] , [0182a]

^Ratio measurements with double fission chambers in the SCK/CEN Cavity Fission

Source [ Gi 84a] . The flux monitor was 239 Pu(n,f) for which the 252Cf fission

spectrum observed value is (1824 ± 1.9%)mb--see Part IA. The 239 Pu fission cross section ratio between 252 Cf and 235U fission spectra is 1.003 based on spectrum evaluation (Section 4.b.3). A conservati ve upper bound for the error in this ratio is ± 0.5%.

5C
TABLE X— 1 7 ( B5 ) . CALCULATED-TO-OBSERVED RATIOS OF CROSS SECTIONS

Spectrum : 235U Thermal -Neutron-Induced Fission

(NBS-Eval uated, [Gr75b], [Gr75c])

Designation : XU5-5-N1

Entry Date : March, 1982

Cross Section : ENDF/BV Dosimetry File Revi sed : July, 1985

REACTION

Observed

(Table X-16)

Cal cul ated

NBS-eval^ 3 ) ENDF/B-V

fiss. spec. fiss. spec.

TABLE X- 1 3 ( B5 ) TABLE X-14a(B5)

Cal cul ated-to-Observed

NBS-eval.( b

) ENDF/B-V

fiss. spec. fiss. spec

(mb) (mb) (mb)

Fission Readitions

239Pu(n,f) [1818 + 1 .9% ] 1786 + 0.2% 1791 [0.982 + 1.9% 0.985]

235 U(n,f) 1200 + 1.9% 1236 + 0.2% 1236 1.030 + 1.9% 1.030

233U(n,f

)

1949 + 3.1% 1912 + 0.2% 1907 0.981 + 3.1% 0.978

238 U( n ,f

)

312 + 2.3% 294.7 + 1.5% 305.2 0.945 + 2.7% 0.978

237Np(n,f

)

1359 + 2.1% 1322 ± 1.0% 1347 0.973 + 2.3% 0.991

232Th(n,f) 82.9 + 3.1% 72.4 ± 1.6% 75.0 0.873 + 3.5% 0.905

240 Pu(n,f

)

1332 + 2.1% 1326 ± 1.0% 1352 0.995 + 2.3% 1.015

He! i urn Production

1 °B( n ,He) 540 + 4.0% 499 ± 2.3% 491 0.923 + 4.6% 0.909

6 Li (n,He) 455 + 4.0% 465 ± 2.0% 455 1.021 + 4.5% 1.000

Threshol d Reactions

93Nb(n ,n '

)

164 + 9 % 154.3 ± 1.4% 159.7 0.94 + 9.1% 0.97

115 In(n,n'

)

190 + 2.1% 173 ± 1.4% 179.3 0.911 2.5% 0.944

4 7T i ( n , p

)

18.0 + 3.5% 21.6 ± 2.2% 22.5 1.200 4.1% 1.250

58 Ni (n,p) 111. + 2.4% 100.9 ± 2.6% 105.0 0.909 + 3.5% 0.946

54Fe(n,p) 81.7 + 2.7% 77.8 ± 2.7% 81.0 0.952 + 3.8% 0.991

46Ti(n,p) 11.8 + 3.5% 10.8 ± 3.8% 11.2 0.915 5.2% 0.949

63Cu(n,a) 0.60 + 7 % 0.540 ± 3.4% 0.558 0.90 ± 7.8% 0.93

48Ti (n,p) 0.307 + 3.5% 0.273 ± 3.3% 0.282 0.889 ± 4.8% 0.958

27At(n, a) 0.720 + 3.5% 0.693 ± 3.6% 0.719 0.963 ± 5.0% 0.999

(^Errors correspond to 235U fission spectrum uncertainties (lo) given in Table *-• and propagated according to the second term of Eq. (X-7), Section 5. a. 3., Part

(^Errors are quadrature sum of errors in columns 2 and 3.

+ See footnote in Table X-16.

67
TABLE X- 18 . OBSERVED SPECTRAL INDEXES

Spectrum : 235U Thermal -Neutron-Induced Fission Desi gnation: XU5

Entry Date : May, 1978

a/U8 -lobs = [Wu238 )]obs Revised : July, 1985

Reaction^ Reaction^)

Non-overl ap

Interval (MeV)

Spectral ) Index

Measurement

Reference

Threshold Reactions

237 Np(n,f)/U8

+ 0.67 - 1.5 4.36 ± 1.7% [Gi 84a]

240 Pu( n ,f )/U8

+ 0.67 - 1.5 4.27 ± 1.7% [Gi84a]

93 Nb(n,n' )U8 0.99 - 1.5 0.526 ± 9 % [A1 82a]

1 15 In( n ,n ' )/U8 1.2 - 1.5 0.609 ± 2.0% [Gi84a], [Fa75a]

47 Ti(n,p)/U8 1.9 - 1.5 0.0577 ± 4.2% [Ma84b], [ K i 78a]

58 Ni (n ,p)/U8 2.1 - 1.5 0.357 ± 2.0% [Gi 84a] , [La82a]

,

[Fa75a]

54 Fe(n ,p)/U8 2.5 - 1.5 0.262 ± 3.5% [Ma84b] , [La82a]

46 Ti ( n ,p)/U8 3.8 - 1.5 0.0378 ± 3.0% [Ma84b]

63 Cu(n,a)/U8 4.7 - 1.5 1.92 E-3±10 % [Fa75a], [ Ki 78a]

48 Ti ( n ,p)/U8 5.9 - 1.5 0.984E-3± 4.2% [Ma84b], [ Ki 78a]

27 A£(n,a)/U8 6.5 - 1.5 2.31 E-3± 4.2% [Ma84b], [ Fa75a]

Non-Threshold Reactions

239 Pu( n ,f )/U8

+ 0.26 - 1.5 5.83 ± 1.7% [Gi 84a]

235U( n ,f )/U8

f

0.20 - 1.5 3.85 ± 1.7% [Gi 84a]

233 U( n ,f )/U8

+ 0.21 - 1.5 6.25 ± 3.0% [Gi 84a]

10 B(n,He)/U8 0.07 - 1.5 1.731 ± 4.5% [Gr85a] , [0182a]

6 Li (n,He)/U8 0.17 - 1.5 1.459 ± 4.5% [Gr85a] , [0182a]

^ a ^All spectral indexes are taken relative to the 238 U(n,f) reaction labeled

" 1) 8

".

^^Non-overlapping energy response interval between the 95% response energy

boundary for each detector. See Table X- 13 ( B5 ) and footnote (b).

^Fission detector indexes are from simultaneous measurement of fission rates

from both isotopes and hence show smaller uncertainties because of reduced

neutron fl uence errors. Other spectral indexes are formed from independent

cross section measurements listed in Table X- 16

.

t Ratio measurements obtained with NBS double fission chamber at the SCK/CEN

cavity fission source.

68
TABLE X-19(B5) . CALCULATED- TO-OBSERVED RATIOS OF SPECTRAL INDEXES - NBS EVALUATION

Spectrum : 235U Thermal -Neutron-Induced Fission

(NBS evaluation, [Gr75b], [Gr75c])

Spectral Index : Calculated: Table X-14(B5)

Observed: Table X- 1

8

Desi gnati on : XU5-5N1

Entry Date : May, 1978

Revised : July, 1985

C a/U8 " [ S a/U8 -leal C . 1 [ S a/U8^obs. 0r C U8/a

Reaction^ Reaction^) Cal cul ated- E R R 0 R S (la)

(c) Non-overl ap To-Observed Observed Cal cul ated Total

Interval (MeV) Val ues Val ue Val ue

Threshold Reactions

U8/ 237 Np(n,f

)

0.67 - 1.5 0.971 ± 1.7% ± 1.0% ± 2.0%

U8/ 93 Nb(n,n' ) 0.99 - 1.5 1.01 ± 9 % - - U8/ 1 15 In( n,n '

)

1.2 - 1.5 1.036 ± 2.0% ± 0.3% ± 2.0%

47Ti (n,p)/U8 1.5 - 1.9 1.27 ± 4.2

%

± 1.1% ± 4.3%

58 Ni ( n ,p)/U8 1.5 - 2.1 0.958 ± 2.0% ± 1.7% ± 2.6%

54 Fe(n,p)/U8 1.5 - 2.5 1.008 ± 3.5% ± 1.9% ± 4.0%

46Ti (n,p)/U8 1.5 - 3.8 0.971 ± 3.0% ± 3.1% ± 4.3%

63 Cu ( n , a)/U8 1.5 - 4.7 0.954 ±10 % ± 2.8% ±10.4%

48Ti (n,p)/U8 1.5 - 5.9 0.940 ± 4.2% ± 2.8% ± 5.0%

27 A£(n,a)/U8 1.5 - 6.5 1.018 ± 4.2% ± 3.2% ± 5.3%

Non-Threshold Reactions

U8/ 239 Pu(n,f

)

0.3 - 1.5 1 ow 0.962 ± 1.7% ± 1.4% ± 2.2%

5.2 - 6.7 high

U8/ 235U(n,f) 0.2 - 1.5 1 ow 0.919 ± 1.7% ± 1.7% ± 2.4%

5.2 - 6.7 high

U8/ 10 B(n,He) 0.066- 1.5 1 ow 1.022 ± 4.5% ± 3.4% ± 5.6%

5.2 - 6.7 high

U8/ 6 Li (n,He) 0.17 - 1.5 1 ow 0.925 ± 4.5% ± 3.1% ± 5.5%

5.7 - 6.7 high

(^Spectral indexes are taken with respect to the 238U(n,f) reaction labeled

"U8" with the lower energy response detector in the denominator. The

calcul ated-to-observed ratio for of (U238) is 0.945 (NBS-eval.) from Table

X-17(B5).

^Non-overlapping energy response interval between the 95% response energy

boundary for each detector. For non-threshold reactions, two non- overlapping energy intervals are listed: The 95% exclusion interval is

labeled "low" and the 5% exclusion interval "high". See Table X-13.

(^Errors in the calculated spectral indexes are for 235U fission spectrum

uncertainties only as given in Table X-5 and propagated according to the

second term of Eq. (X-8), Part IA, Sect. 5. a. 3. The error in o

f (U238)

calculated is ± 1.5% from Table X-17(B5).

69
NBS CAVITY FISSION SOURCE

235u SOURCES

DETECTOR STACK

CADMIUM ENCLOSURE

Figure x-3. NBS Cavity Fission Source and detector assembly

70
AR

f

0.1 CM THICK

fas CADMIUM TUBE

. ALUMINUM

CLAD

FISSION

CHAMBER

2 R,

2 Z,

U-235 FOIL DIMENSIONS

SOURCE 2Zo 2 R 0 AR

ASSEMBLY NO. (CM) (CM) (CM)

MARK II 7.7 3.3 0.01

MARK IIA 7.96

"

' | 3.44 0.02

MARK III 7.95 5.25 0.04

\

CYLINDRICAL

U-235 FOIL

Figure X-4. Cavity Fission Source Irradiation Facility at. SC K/ CEN Belgium

71



Part I I A: Intermediate-Energy Standard Neutron Fields (ISNF)

With Thick A °B Shell

73
TABLE OF CONTENTS

Page

1.

IDENTIFICATIONS 75

l.a. Title, Classification

l.b. Designation

l.c. Entry Dates and Revisions

l.d. Generic Description

l.e. Facility Location

l'.f. Contacts for Information

2. SUMMARY INFORMATION 77

3. FACILITY DESCRIPTION 78

3. a. Configuration and Character!' sties

3.b. Irradiation Procedures

(experimental arrangements and fluence monitoring)

3.

c. Specification for Neutron Transport Calculation

(materials and geometry)

4. NEUTRON FIELD CHARACTERIZATION 81

4.

a. Neutron FI uence

(intensity, fluence transfer, gradients)

4.

b. Neutron Spectrum

4.b.l. Calculation

(sensitivity study)

4.b.2. Measurement

4.

b.3. Evaluated Spectrum

5. INTEGRAL DETECTOR RESPONSE 96

5.

a. Calculated Reaction Probabilities

5.

a.l. Spectrum Response and Error Propagation Tables

(cross sections, spectrum fraction, response range,

propagation of spectrum errors)

5. a. 2. Spectral Indexes

5. a. 3. Spectral Index Errors

5.b. Measured Reaction Probabilities

5.c. Measured Cross Sections and Cal cul ated-to-Observed Ratios

6. LIST OF TABLES 101

7. FIGURES 118

74
1. IDENTIFICATIONS

l.a. NEUTRON FIELD : INTERMEDIATE-ENERGY STANDARD NEUTRON FIELD (ISNF) WITH

THICK 10B SHELL

CLASSIFICATION : STANDARD NEUTRON FIELD

l.b. DESIGNATION : ISNF(5)-1-L1

Discrete ordinates transport calculations (LASL 150-group

ONEDANT) with ENDF/B-V boron and carbon cross sections

and with the fission source spectrum taken as the average

of the ENDF/B-V and NBS evaluations. References:

[Br79a] , [Ei77a], [La80a], [So79a]

l.c. ENTRY DATE : July, 1983 REVISIONS : July, 1985

l.d. GENERIC DESCRIPTION :

The Intermediate-Energy Neutron Field (ISNF) is an irradiation facility

designed to produce a strong component of neutrons in the energy range of

interest for fast-reactor and related technologies. Arranged in spherical

geometry, driven by fission neutrons, and employing two materials with

uniquely wel 1 -understood nuclear properties for neutron transport, the

facility provides a constant flux of neutrons whose spectrum may be accurately

calculated. General properties of the neutron field at the center of ISNF

as follows:

Neutron fluence rate 1.5 x io 9 n/cm 2 s

Irradiation space 5 cm dia. sphere

75
Spectrum :

Energy range 98% of fl uence between 1.2 keV and 5.7 MeV

Median energy 0.58 MeV

Median response energy for reaction with 1/v cross section 25 keV

Neutron transport in ISNF is determined almost entirely by fission neutron

scattering and diffusion in carbon, and by neutron absorption in 10 B.

The resulting neutron spectrum, established on the basis of neutron transport

calculations in 1-D geometry, is limited in accuracy only by the uncertainty

in the driving spectrum and the input cross section data. Such calculations,

employing both di screte-ordi nates and Monte-Carlo methods, have been carried

out by NBS [Ei77a], Los Alamos [La80a], [So79a], and Oak Ridge [Wa81a]. The

simplicity of a 1-D calculation permits detailed investigation of spectrum

sensitivity to uncertainties in physical and nuclear parameters of the system

(mainly material densities and cross sections of carbon and 10B), and spectrum

perturbations due to extraneous materials required for fabrication.

The neutron fl uence in ISNF has been established to an accuracy of ± 2.1% by

means of neutron fluence transfer from the 252Cf fission neutron standard

field [Gr75c], [Gr77b], [La84a]. Fission cross sections for most fissionable

isotopes of interest for nuclear technology have been measured in the facility

with fission ionization chambers; other measurements include capture cross

sections in gold and 238U, and He-producti on in 10B and 6Li

.

76
l.e. FACILITY LOCATION:

National Bureau of Standards

Center for Radiation Research

Gaithersburg, MD 20899

Contact: Dave Gilliam or James Grundl

Phone: 301-921-2767

l.f. CONTACT FOR INFORMATION :

Charles Eisenhauer or James Grundl

National Bureau of Standards

Center for Radiation Research

Gaithersburg, MD 20899

Phone: 301-921-2767

2. SUMMARY INFORMATION

2. a. AVAILABLE NEUTRON EXPOSURES

• fluence rate

® nominal maximum fluence

• accuracy of free-field fluence (la)

• irradiation space

1.5 x 10 9 n/cm 2

s

1 x 10 14 n/cm 2

± 2 . 1 %

5 cm dia. sphere

2.b. DETECTOR RESPONSE RANGE AND UNCERTAINTY

E 50 (median energy):

235U( n,f ) 237Np( n ,f ) Au( n , y)

Energy response range (Table 1-5):

0.26 MeV 1.54 MeV 6.5 keV

0.90 keV 0.53 MeV 0.33 keV

3.1 MeV 4.5 MeV 0.97 MeV

Cross section error associated with spectrum uncertainties (Table I-

± 0 . 6 % ± 1 . 2 % ± 2 . 0 ?

95 -

05 '

77
3. FACILITY DESCRIPTION

The Intermediate-Energy Standard Neutron Field (ISNF) operates in the graphite

thermal column of the Research Reactor at the National Bureau of Standards in

Gaithersburg, Maryland. As an irradiation facility it is in general use for

neutron cross section measurements and detector calibrations. As a primary

neutron field standard its physical components and geometric configuration

have been extensively investigated and it is these features of the system

which are summarized in this section.

3. a. CONFIGURATION AND CHARACTERISTICS [Ei77a], [Gr77c], [Mc82a], [Wa81a]

With the reactor at full power, the ISNF facility is assembled for operation

into a 30 cm x 30 cm thermal column opening. The position of the ISNF cavity

inside the reactor thermal column (a 1.4 x 1.3 x 0.94 meter graphite block) is

shown in Fig. 1-1. A schematic cross section of the ISNF arrangement within

the cavity is shown in Fig. 1-2. Three levels of access to the irradiation

region are available: (1) instruments and irradiation samples may be inserted

or removed through a 5 cm diameter cylindrical channel; (2) the 10B shell and

graphite pieces to which it is attached slide in a 20 cm diameter cylindrical

access; and (3) the entire cavity block may be withdrawn. Eight 235U source

disks are mounted symmetrically near the surface of the cavity of which four

are shown in Fig. 1-2.

The flanged cadmium shield fits over the stems of instruments placed inside of

the shell and prevents thermal neutrons from streaming through the hole in the

shell access plug. Shell access plugs without a hole are used for passive

detector irradiations which do not involve instrument stems. Streaming

78
through access holes is a characteristic problem for any driven, onedimensional neutron field. Experimental investigation is required to demonstrate that thermal and wall -return neutron streaming is a negligible

perturbati on.

The critical component of ISNF is the 10B shell which establishes the spectrum

shape at energies below about 180 keV. The shell, formed as stepped

hemispheres, at the Los Alamos National Laboratory by techniques of powder

metallurgy is 14.4 cm outside diameter and 1.29 cm thick with a total mass of

1389 g and a mean 10B thickness of 1.28 g/cm 2. The powder used for fabrication of the shell was a mixture of 92.7% enriched 10B metal and aluminum.

The fission source disks of 93% enriched uranium metal, 16 mm dia. x 0.15 mm

thick, are mounted 1 cm from the cavity surface. The thermal neutron fluence

rate in the cavity is ~ 2 x 10 11 n/(cm 2 s) and each disk operates at two watts

of fission power. The total fission neutron source strength for the eight

disks is 1 .1 x io 12 s * 1

.

3.b. IRRADIATION PROCEDURES

Irradiation experiments are placed inside of the 10B shell with the cavity

block outside of the reactor. The block is then inserted into the thermal

column and irradiations are initiated by raising the boral curtain behind tne

thermal column. At the end of the irradiation the curtain is lowered and

after a suitable cool-down time the block is removed. Alternatively, experiments can be removed by withdrawing the 10B shell assembly which is attached

to the shell access plug. This procedure leaves the highly radioactive

fission disks inside of the thermal column.

79
The central region of known spectrum and minimal fl uence gradients extends out

to a radius of ~ 2.5 cm as shown by the dashed circle in Fig. 1-2. Neutron

fl uence monitoring is by means of an activation monitor foil or a fission

chamber.

3.c. SPECIFICATION FOR NEUTRON TRANSPORT CALCULATION [Ei 77a], [Gr77c],

[Mc82a] , [Wa81a]

The ISNF is a spherically symmetric system of 10B and carbon with some

unavoidable amounts of nB and aluminum and a minute amount of moisture. The

fission source disks which drive the system are mounted 1 cm from the cavity

wall. Because of spherical symmetry, the fluence contribution at the center

from each disk depends only on its source strength and distance from the

center, not on its angular position; hence, for the neutron fluence at the

center, the eight fission disks considered separately or as a set are

equivalent to a shell source.

Material densities and dimensional specifications for purposes of transport

calculation are given in Fig. 1-3. The two zones shown in the graphite are a

one-dimensional representati on of the measured density difference between

thermal column graphite and the cavity insert blocks — see Fig. 1-1. The

hydrogen density given for the graphite corresponds to 0.22 weight percent of

moisture — see Section 4 . b . 1 . The only significant compromise with the real

ISNF system in the calculations is the substitution of carbon for UB in the

boron shell. For discrete ordinates, the following computational structure

provides sufficient accuracy:

80
Angular quadrature S

8 Legendre expansion for carbon scatter P

3 Spatial mesh vacuum (0-6 cm) 10 intervals

boron shell (6-7 cm) 8 intervals

vacuum (7 - 15 cm) 9 intervals

graphite (15 - 65 cm) 28 intervals

Spectrum uncertainties associated with errors in densities and dimensions of

components are taken up in Section 4.b.

4. NEUTRON FIELD CHARACTERIZATION

The ISNF neutron spectrum divides naturally into three energy regions, each

dominated by a characteristic feature of the system. For energies above

3.7 MeV, the spectrum resembles that of the 235U fission neutrons which drive

the system; between 0.18 MeV and 3.7 MeV, the spectrum is dominated by

neutrons returning from the graphite after several collisions; the remainder

of the spectrum below 0.18 MeV is determined by 1/v-absorpti on in the 10 B

shell. Partitioned in this way, the ISNF neutron spectrum is as follows:

Three-group characterization

Group

fl uence

235U( n,f

)

response

f racti on

10B(n,a)

response

f racti on

Fission spectrum region (> 3.7 MeV): 0.043 0.031 0.013

Graphite return region (0.18 -3.7 MeV): 0.673 0.52 0.22

10 B absorption region (< 0.18 MeV): 0.284 0.45 0.77

Spectrum boundaries Median energy

E

p (p = 0.99) = 1.2 keV E (P

P = 0.50) = 0.58 MeV

E

p (p = 0.01) = 5.6 MeV

81
Although most of the neutron spectrum lies in the graphite return region, much

of the response of a 1 ow energy detector like 235U fission lies in the 10B

absorption region. This shift is even more pronounced for a 1/v cross section

like 10 B(n,a).

The complete ISNF spectrum (Fig. 1-4) is defined as the 150-group ANISN

calculation performed at Los Alamos. The input data are the geometric and

physical parameters given in Fig. 1-3, the ENDF/B-V cross sections, and for

the driving source, an average of the ENDF/B-V and NBS evaluated 235U fission

neutron spectrum shapes. The angular quadrature (S

8

) and the spatial mesh for

this calculation were shown to be satisfactory in early exploratory computations [Fa74b]. A coarse 14-group energy structure is employed for summarizing

the detailed spectrum, for comparisons with other calculations, and for

stating spectrum accuracy. The defined central spectrum for ISNF with a thick

shell in the 14-group energy structure is given in Table 1-1 along with

results from three validating calculations.

The large gamma ray intensity in the ISNF cavity comes primarily from thermal

neutron capture in graphite. The neutron-to-gamma ratio is about

2 x 10 4 (n/cm 2

s )/ (R/h)

.

4. a. NEUTRON FLUENCE

The ISNF is driven by eight 235U fission disks which yield a central neutron

fl uence rate of ~ 1.5 x 10 9 n/cm 2 s, and for a nominal maximum irradiation

time of 20 hours, deliver a neutron fl uence of 1.1 x 10 14 n/cm 2

.

82
The central ISNF fluence is established by means of neutron fl uence transfer

from the 252Cf Fission Neutron Irradiation Facility (see Section 3.b of

Part IB and Refs. [Gr78a], [La84a]). The fluence transfer reaction is

239 Pu(n,f) and the fluence transfer expression is

aCf

(>E

p

) ^ %f (>E

p

)

_ °ISNF . f n

ISNF aISNF ( >E

p

) ^ISNF ^ >E

p

^ °Cf

Cf

where $ = total neutron fluence, and the subscript "ISNF" denotes the central

ISNF fluence at the 239Pu(n,f) monitor position and "Cf" the 252Cf

fission neutron fluence associated with the calibration of the

239Pu(n,f) detector.

The ratios on the right side of Eq. (1-1) are, in order, 239Pu(n,f) truncated

fission cross sections a(> E^), spectrum fractions ^(> E

p

), and the ratio of

the 239Pu(n,f) reaction rate obtained in the ISNF to that obtained at the

252Cf irradiation facility— see Section 5. a in Part IA. For 239Pu(n,f) the

truncated cross section ratio is 1.031 and the spectrum fraction ratio is

0.952. The error in the product, 1.033 * 0.952 = 0.984, attributable to 252Cf

and ISNF spectrum uncertainties is less than 0.3%. However, the 239Pu(n,f)

cross section shape error including covariance adds an additional uncertainty

of ± 1.6%. Combined with a typical error of ± 0.8% for the reaction rate

ratio, and ± 1.2% for the 252Cf fission neutron fluence <J>^, the fluence

transfer procedure yields a free-field central ISNF neutron fluence with an

accuracy of ± 2.1% (la).

83
The central ISNF fl uence rate also has been estimated on the basis of a

cursory determination of the neutron source strength of all eight fission

disks. This is an involved measurement not amenable to high accuracy. Its

purpose was to provide a gross check of the total neutron return fl uence in

the transport calculation. The assigned ISNF calculation (see Section

4.b.l.(a)) gives a value of 1.36 x 10“ 3 n/cm

2

s for the central fl uence rate

per source neutron for this parameter. The measured value is

1.27 ± 0.07 x lO3

.

Fl uence variation over ± 1.5 cm of the central region of the ISNF field is

less than ± 3%. There is some asymmetry in this variation about the geometric

center due to unequal source strengths among the fission disks. This slight

gradient does not create any spectrum change, and for conventional irradiation

experiments carried out with disk-shaped detectors, one of which is a fl uence

monitor, the uncertainty in the spatial averaged fl uence is negligible.

4.b. NEUTRON SPECTRUM

The fundamental design concept behind ISNF was to create a family of neutron

fields whose neutron spectrum could be established entirely on the basis of

neutron transport calculation. The following features of the ISNF system were

designed to match this concept: (1) simple one dimensional (spherical)

geometry; (2) restriction to materials for which neutron cross sections are

among the best known; (3) 235U fission spectrum as the driving source; and

(4) spatial decoupling of the elements which dominate the neutron transport

problem (i.e., separation of the neutron source, moderator, and absorber).

The arrangement shown in Fig. 1-2 consists of a spherical cavity in a large

block of graphite, a shell of 10B lightly supported at the cavity center, and

84
thin fission source disks of 235 U metal placed symmetrically around the

periphery of the cavity. Fission neutrons generated in the source disks, both

the uncollided and those returning from the graphite, are transmitted by the

i0 B shell and give rise to the ISNF field at the center of the cavity. Proper

convergence of discrete ordinates calculations for the ISNF geometry and

materials is discussed in Refs. [Fa74b] and [Br80a].

The neutron transport problem is dominated by scattering in carbon and absorption in i0 B. Therefore, as far as nuclear data is concerned, the neutron

spectrum is determined chiefly by the 235 U fission neutron spectrum (above

~ 3 MeV only), by the kinematics of elastic scattering in carbon, and by two

accurately known cross sections: a. . for carbon, and a . for i0 B. Energy

dependence only is required for i0 B absorption because neutron transmission

thru the boron shell is determined experimentally in a 2 keV monoenergeti

c

neutron beam. Spectrum calculations in spherical geometry are simple and

exact, limited in accuracy only by uncertainties in the input fission spectrum

and cross section data. Because of this simplicity, it is easy to investigate

the sensitivity of the spectrum to uncertainties in the physical and nuclear

parameters of the ISNF system, as well as to scattering in extraneous

structural materials.

In the remainder of this section neutron transport calculations will be

summarized, an evaluated ISNF spectrum will be described, and spectrum sensitivities to physical and nuclear parameters will be established for various

types of detector responses. The evaluated ISNF spectrum is given in SesMon

4.b.3 along with errors based on a composition of the spectrum sensitivity

studies and a quantitative estimate of ISNF parameter uncertainties.

85
4.b.l. Cal cul ation . Five neutron transport calculations make up the

cal cul ational base for ISNF. Of these, the 150-group cal cul ati onal result

from Los Alamos National Laboratory was chosen for the assigned spectrum. In

brief outline the five calculations are as follows:

Assigned

(a) Discrete ordinates calculation performed at LANL using ONEDANT in the

P

3 S8 approximation with a 150-group energy structure and ENDF/B-V cross

sections. The angular quadrature (S

8 ), the spatial mesh (73 space intervals from the center out to 50 cm into the graphite), and the outer

graphite boundary have been investigated for convergence. The assigned

spectrum is taken as the average of two calculations, one of which uses

the NBS evaluated fission spectrum and the other, the ENDF/B-V fission

spectrum shape. The designation I SNF ( 5

) -1 -LI refers specifically to this

calculation. These calculations are entirely similar to those carried

out earlier with ONETRAN (see (e) below) and references for the latter

remain applicable [So79b], [La80a].

Suppl ementary

(b) Discrete ordinates calculations performed by NBS using ANISNW in the

P

3 S8 approximation with a 40-group energy structure [Gr77c], Groupaveraged cross sections were prepared from ENDF/B-III using SUPERTOG and

the GAM- 1 1 energy structure [Fa74b]. This calculation is the reference

for most of the ISNF sensitivity studies.

(c) Discrete ordinates calculation performed by ORNL using XSDRNPM in the

P

3 S8 approximation, ENDF/B-IV cross sections, and the 171-group VITAMIN-C

energy structure [Wa81a], [Br80a], [Br79b], [Gr76a].

86
(d) Monte-Carlo calculations performed with ENDF/B-IV cross sections by

LANL using MCNP in two options: (i) discrete-reaction and continuousenergy kinematics, and (ii) complete continuous energy [La80a], [So79b],

[LA78a]

.

(e) Early LANL calculations with ONETRAN, 240 energy groups, and the NBS

evaluated fission spectrum. Otherwise similar to (a) above [So79b],

[Hi 75a], [La80a],

The assigned spectrum normalized to unit fluence is shown complete in Fig. 1-4

and in a 14-group energy structure in Table 1-1. Comparisons with the supplementary calculations are also given in the Table and show departures of more

than 3% in some of the energy groups. This may be attributed to the coarse

group structure of the NBS calculation and to statistical fluctuations in the

Monte Carlo calculation. The comparison of the NBS 40-group calculation with

the LANL 150-group illustrates the general insensitivity (except for the

lowest energy groups) of spectrum calculations to energy group structure.

Two components of the ISNF spectrum are physically realizable and may be

distinguished because of spatial decoupling of the neutron source, the neutron

moderator, and the neutron absorber. The first component is the 2 35|

J fission

neutron source which drives the system and the second is the neutron return

from the graphite plus the fission source neutrons. The second component

exists in the ISNF cavity when the boron shell is removed. This empty cavity

spectrum relaxes smoothly into a near-l/E spectrum characteristic of a pure

scattering medium and as such is also a standard neutron field. Both

components are shown in Fig. 1-5 along with the composite central ISNF

87
spectrum which results from the transmission of the second component through

the 10B shell. The extent and significance of spectrum decoupling in ISNF is

discussed in Ref. [Fa74b],

Sensitivity Studies . The accuracy of the ISNF spectrum is assessed by means

of a sensitivity study involving the variation of physical and geometric

parameters, input nuclear data for boron and carbon, and the fission spectrum.

Using the 40-group NBS calculation, sensitivity factors are derived for the

spectrum itself and for selected spectrum-averaged reaction cross sections.

For integral reaction rate measurements, the error associated with the ISNF

spectrum is the quadrature sum of the cross section sensitivity factors

correspondi ng to each ISNF parameter uncertainty.

The parameters studied are as follows:

1. Absorption in the 10B shell

[shell thickness (at/b)] x [ 10 B cross section (b/at)]

2. Mean-f ree-path in graphite:

[graphite density x carbon total cross section] -1

3. Anisotropy of carbon scattering

4. Position of source disks

5. Fission spectrum

6. Scattering in the 10B shell:

(a) modulation at aluminum resonances; (b) total scatter in

27Aj£, 1 : B , and impurities; (c) high-energy scattering in 10 B

7. Scattering in aluminum support structures within the cavity.

88
(1) Absorption in 10B shell . The key parameter for determining the shape of

the ISNF spectrum below 100 keV is the absorption thickness of 10B in the

shell. Uncertainties in the 10B cross section a(b/at) or the mass thickness

X ( at/b) are amplified in the spectrum at low energies because of the 1/v

behavior of the absorption cross section. The 10 B mass thickness for the

shell is determined from a measured 10B enrichment, the mass of the finished

shell, and a chemical analysis of the boron-aluminum powder mix employed. The

neutron transmission of the finished shell, measured with a 2 keV filtered

neutron beam provides an independent check of the absorption thickness of the

shell. An observed shell transmission of 0.281 was found to be uniform over

accessible portions of the shell to better than 1%. The 10B mass thickness

deduced from the transmission measurement, using ENDF/B-V cross sections for

10 B, is (3.9 ± 2 )% greater than the thickness obtained from chemical and

isotopic analysis and physical weighing. An average of these two values was

used for the transport calculations.

The observed neutron transmission is important because it puts into the transport calculation a measured value of aX, the neutron absorption in the shell.

With the shell transmission established by experiment at one energy, the lowenergy side of the spectrum depends only upon the 1/v behavior of the 10 B

absorption which is a cross section standard below 100 keV [Ca84a]. This

behavior has been established both by measurement and by theory. From thermal

to 1 keV it is accurate to ± 1%; up to 10 keV it is within ± 2%, and up to

100 keV within ± 3% [Ca84a], [Wa77a], [Ha76a],

The uncertainty in aX for the shell has been set equal to ± 2%, the depart, jr.-

of the two mass thickness determinations from the average value. The

89
change associated with a 2% increase in 10 B thickness is given in the third

column of Table 1-2. Below about 1 keV the uncertainty in 10 B thickness is

seen to dominate the ISNF spectrum error; above 20 keV it is unimportant.

(2) Mean-f ree-path in graphite (1/po) . Measured values of graphite density

exist for various unused thermal column stringers (standard deviation of

± 0.3% among the densities of four individual pieces), and for two of the

cavity assembly pieces. These two sets of densities, stringers and cavity

blocks, differ by 4.5% and are modeled into the one-dimensional calculation.

The spectrum difference between the two-zone density model and uniform average

graphite density is very small. The total cross section for carbon has been

studied extensively and its error is below ± 0.75% in the energy region below

0.1 MeV and ± 1% above 0.1 MeV [He75b], Based on the density measurements and

the cross section data, the uncertainty in the graphite mean-f ree-path is set

conservatively at ± 2% for all energies. The spectrum change associated with

a 2% decrease in mean-f ree-path is given in the fourth column of Table 1-2.

The graphite stringers were also analyzed for moisture by the Karl Fischer

titration method. The resulting hydrogen to carbon atom ratio in the graphite

from this measurement is 0.003 ± .0003. Calculated spectrum differences are

listed in column 8 of Table 1-2 and corresponds to an estimated error of ± 50%

for the in situ moisture content of the graphite in the thermal column. The

differences are less than 0.5% except in the low keV range where they reach

0 . 8%.

(3) Anisotropy of carbon scattering . Angular distributions of scattering in

carbon have been studied in great detail over the energy range of the ISNF

90
spectrum and the results have been interpreted by multilevel R-matrix

analysis. These investigations make the carbon scattering cross section a

reference standard in many experimental studies of neutron interactions

[Sm78a], Angular distributions in discrete ordinates transport calculations

are described by Legendre harmonic expansions truncated at P . The effect of

scattering anisotropy was examined by repeating the calculation with truncation at P

3 reduced to PQ

, P

x

, and P

2

. Pure isotropic scattering, the P

Q approximation, differs from the P

3 approximation by more than 10% in the MeV

range of the spectrum and up to 3% at other energies. The P

1 component

provides some improvement but only with the P

2 approximation do the departures

from the P

3 calculation become less than 1.5% at all energies. The agreement

of Monte Carlo calculations with the discrete ordinates calculations (see

Table 1-1) verifies that scattering in graphite is treated adequately by

harmonic expansion in the discrete ordinates computation. The spectrum error

from anisotropic scattering in carbon is taken as the spectrum change between

the P

2 and P3 approximati on . The associated spectrum variation is given in

column 7 of Table 1-2.

(4) Position of source disk . The enriched uranium metal source disks

(0.5 g each) are mounted in thin-walled aluminum holders which are press

fitted into the graphite wall of the cavity. The uncollided fission neutrons

from these disks constitute the largest part of the central fluence above

3 MeV. Hence, there is an inverse-r 2 dependence of the central fast fluence

on the position of the sources. The effective source disk position uncertainty is set conservatively at ± 1 mm. The spectrum change associated wit* a

decrease of 1 mm in source radius is given in column 6 of Table 1-2.

91
(5) Fission spectrum Fission neutrons dominate the ISNF spectrum above

3.7 MeV so the error in this energy range is taken to be that of the 235U

fission spectrum. Uncorrelated errors for the NBS fission spectrum evaluation

are given in a seven-group energy structure in Table X-5 of Part IA. In order

to examine the sensitivity of the entire ISNF spectrum to the fission spectrum

it is necessary to have an alternative fission spectrum shape which departs

credibly from the NBS evaluation. The fission spectrum shape recommended in

the ENDF/B-V file was chosen as an acceptable alternative because the difference is similar to the errors given in Table X-5. A comparison of the NBS and

ENDF/B-V evaluations is shown in Table 1-3. Separate calculations of ISNF

with the two source spectra provide a basis for estimating the ISNF spectrum

error attributable to fission spectrum uncertainties.

Time-of-fl i ght measurements have established the average energy of the 235U

spectrum relative to that of 252Cf [Sm79a], Because the 252Cf spectrum is

better known, as indicated by the reduced errors in Table X-5, and because the

reported difference in average energies in Ref. [Sm79a] agrees better with the

NBS evaluation than with the ENDF/B-V value, half the departure of ISNF

calculations obtained with these two fission spectra as the neutron source

were judged appropriate as a source spectrum error. The spectrum change

associated with substituting Xgyfor x^gg i n the ISNF calculation (divided by

2) is given in column 4 of Table 1-4.

(6) Scattering in boron shell .

(6. a) Total scatter in 27At, nB, and other impurities produce small

perturbations in the spectrum. Aluminum, at 26 atom percent of the base

material, is the most abundant extraneous material in the shell. An

92
uncertainty of ± 10% in the transmission coefficient associated with this

amount of aluminum leads to modest spectrum changes in terms of a multi group

description (see column 5 of Table 1-2). Localized modulation of the spectrum

at aluminum resonances is discussed below. The 11B content at 3 atom percent

and other impurities (C, Si, Fe) are all entered into the calculation as

carbon at 4.5 atom percent. The spectrum uncertainty attributable to these

minor constituents is covered by the uncertainty assigned to the aluminum.

( 6 .b) Modulation of the ISNF spectrum near an aluminum resonance is a

striking feature of the ISNF spectrum as displayed in Fig. 1-4. This effect,

beginning with the resonance at 34 keV, is the result of a redistribution of

elastically scattered neutrons in a localized energy region. In support of

this i nterpretation, the modulation near an aluminum resonance was reproduced

by means of an elementary single scattering analysis involving an aluminum

shell of equivalent atoms/cm 2 placed just inside the boron shell. The

scattering is treated as resonance-energy removal of neutrons directed

radially inward, and inscatter of the remainder with a spectrum degraded in

energy by the average logarithmic decrement for elastic scatter (5 = 0.072).

The expression obtained with this analysis is

-ME')t

where E' = (1 + ?)E.

ISNF w/A

£

= . ISNF w/o Ml 1 1 - e -E

J

(E)t

Figure 1-6 shows the results of this single-scatter analysis for the prominent

resonance at 34 keV. The smooth ISNF spectrum without aluminum, the negative

93
removal contribution, and the positive inscatter contribution are shown

separately. The resulting spectrum accurately represents the spectrum

calculated by the discrete ordinates method. The difference, indicated by

cross hatching in Fig. 1-6, is only about 1% of the neutron fl uence between 20

keV and 50 keV or about 0.065% of the total ISNF fl uence. The seven

resonances in aluminum which may be distinguished in the ISNF spectrum are at

5.9, 34.7, 88.5, 119.8, 145, 159 and 205 keV. The effect of this spectrum

modulation upon integral reaction rates is small, and the associated error

much less, even if there are matching resonances.

(6.c) Scattering in 10B above 0.1 MeV has little affect as may be seen

by comparing ISNF spectra with and without shell in Fig. 1-5.

(7) Scattering in aluminum . The support structures within the cavity not

included in the specification of ISNF are small. The total mass of the tubes

and rings involved is about 40 grams or 6% of the amount of aluminum in the

boron shel 1 .

4.b.2. Measurement . The ISNF is a standard neutron field designed to be

characterized by one-dimensional neutron transport calculations. Only in the

energy range above 1 MeV, where the ISNF is dominated by uncollided fission

neutrons, does the spectrum depend upon measurement, namely the fission

spectrum itself — see Section 4.b., Part IA. The errors associated with

calculation as discussed in the previous section, are smaller than those

involved with any type of differential neutron spectrometry . This is

particularly true for the spectrum below 50 keV, the definition of which is

the most distinctive feature of the ISNF system.

94
4.b . 3. Evaluated Spectrum. The assigned spectrum for ISNF is the

average of two Los Alamos 150-group ONEDANT calculations which employ two

different 235 U fission spectrum shapes: the ENDF/B-V form and that of the NBS

evaluation. A summary of the assigned ISNF spectrum in 14-groups is given in

Table 1-4 and a graphical display in Figure 1-4. The complete 150-group

spectrum is available on request.

Spectrum changes associated with errors in all significant nuclear and

physical parameters, as discussed in Section 4.b.l, are displayed in

Table 1-2. The changes are less than 1% except at extremes of the spectrum.

The quadrature sum of these parameter-associ ated errors for each energy group

characteri zes the overall accuracy of the spectrum. They are listed in

column 3 of Table 1-4. The remaining spectrum uncertainty associated with the

235 U fission spectrum source is listed in column 4 of Table 1-4 — see

Secti on 4.b .1 ( 5)

.

The composite group fluence errors set out in Table 1-4 indicate a la bound

within which the true ISNF spectrum is estimated to be. For example, the lo

error bound in the energy group 9.12 to 24.8 keV is ± /( 1 . 14

)

2 + l 2 or ± 1.5%.

Such composite errors can aid in establishing the accuracy of a differential

spectrometer calibration in ISNF, but they cannot be used to obtain

uncertainties in calculated reaction probabilities. In the latter case it is

necessary to use the spectrum variations given in Table 1-2 because spectrum

and parameter changes are fully correlated. The change in reaction cross

section associated with each ISNF parameter uncertainty must be established

first and then summed in quadrature to obtain a correct reaction rate errm.

These propagated errors for calculated reaction rates and reaction rate ratir

are treated in Section 5. a.

95
5. INTEGRAL DETECTOR RESPONSE

Neutron dosimetry measurements with integral detectors generally benefit from

the use of benchmark neutron fields for calibrating measurement techniques and

for referencing data interpretation methods. In order to carry out these

measurement assurance procedures, and in particular for estimating

uncertainties, it is necessary to distinguish between calculated and measured

reaction probabilities and to identify response parameters which characterize

the energy response of each detector in the various spectra to which it is

exposed. A formulation which meets these requirements is outlined in

Section 5 of Part IA. Response parameters for a number of integral detectors

including cross sections, spectral indexes, and energy response ranges are

presented in Section 5. a. Calculated reaction cross section errors, including

propagation of spectrum uncertainties, are described in Section 4. Measured

reaction probabilities are given in Section 5.b followed by comparison with

the calculated values in Section 5.c.

5. a. CALCULATED REACTION PROBABILITIES See Part IA, Section 5. a.

5.a.l. Spectrum Response and Error Propagation Tables . Basic integral

detector response parameters for ISNF are given in Table 1-5. The spectrum

averaged cross sections listed in column 2 are the ful 1 -spectrum averaged

values above a cadmium cut-off energy of 0.4 eV and are followed in column 3

by cross sections truncated at a response fraction of p

= 0.95. The spectrum

fraction, [^(> E ) for p

= 0.95], follows in column 4. The energy response

P characteristics of the reaction, given in the last three columns, are the

median response energy and the response range, i.e., the lower- and upperenergy bounds that include 90% of the detector response. The lower-energy

bound, Ep(p = 0.95), is the truncation energy corresponding to the truncated

cross section given in column 3.

96
Each of the ISNF parameter errors discussed in Section 4.b.l corresponds to

spectrum variations listed in Table 1-2. A convolution of these variations

with reaction cross sections in the same multigroup format yields integral

cross section errors attributable to the respective ISNF parameter errors. A

set of such cross section errors for a representative group of reactions is

given in Table 1-6. Cross section changes in ISNF which correspond to the

fission spectrum difference indicated in Table 1-4 (divided by 2) are given in

col umn 3 of Tabl e 1-7.

For fission and threshold reactions, it is appropriate to obtain an ISNF

response relative to that in the 235U fission spectrum. The residual error

for such a cross section ratio in ISNF, attributable to fission spectrum

uncertainties, is the difference between cross section changes associated with

uncertainty in a pure fission spectrum (column 4 of Table 1-4) and cross

section changes in ISNF associated with that same fission spectrum uncertainty

(column 3 of Table 1-7), all divided by two. This net error is given in

column 5 of Table 1-7 for fission and threshold reactions. The latter all

have more than 50% of their response above 0.25 MeV in both spectra.

Total propagated errors for calculated ISNF cross sections are presented in

Table 1-8. The error given in column 4, applicable for absolute cross section

measurement, is the quadrature sum of the total error component from the ISNF

parameters (column 3 of the table) and the fission spectrum error component

from column 3 of Table 1-7. The error in column 5 of Table 1-8 is for cros^

section ratios taken relative to the 235U fission spectrum and is the quadrature sum of the value given in column 3 and that from column 5 of Table 1-7.

Errors for reactions not listed will be similar to reactions in the table wrthe closest median response energy.

97
5. a. 2. Spectral Indexes. (See Section 5. a. 2., Part IA for

formulations)

A selected set of spectral indexes calculated for ful 1 -spectrum-averaged cross

sections from Table 1-5 is given in Table 1-9; a correspondi ng set for

truncated cross sections is listed in Table I - 10

.

5. a. 3. Spectral Index Errors . The cross section errors associated with

individual ISNF parameters given in Table 1-6 are the starting point for

spectral index error propagation. The differences between the errors

corresponding to the two reactions of the index from Table 1-6 are summed in

quadrature and then added to the corresponding error difference from Table 1-7

(column 5 if the index is measured relative to the fission spectrum, column 3

if it is not). The summation for the spectral index 235 U(n,f)/ 238U(n,f) is

set out below as an example:

S 2 = [

( -0.5-0.

2

) 2 + ( 0 .2+0.4 ) 2 - (-0.2+0.

5

) 2 + (-0.1-0.5 ) 2 + (-0.1-0.4) 2

] B-10 X a Source Moisture

c sc

1 — ISNF parameters. Table 1-6 1

+ [(-0.2 - 1 .7) 2 or (-0.2 - 0 .2) 2

]

1 fission spectrum 1

I relative to X25 (Table 1-7, col. 5) —— absolute (Table 1-7, col. 3)

= [1.55] + [3.6 or_ 0.16]

6 = 2 . 3% without referencing to the fission spectrum

= 1.3% relative to the 235U fission spectrum.

98
Errors for representati ve spectral indexes are given in Table 1-11. They vary

from less than ± 0.5% to ± 3% depending upon the separation of the spectrum

response range of the two detectors and whether the spectral index is taken

relative to the fission spectrum. The spectrum associated error combined with

experimental uncertainties establish the total error for a test of cross

sections by means of a spectral index measurement.

5.b. MEASURED REACTION PROBABILITIES (See Section 5.b, Part IA)

5.c. MEASURED CROSS SECTIONS AND CALCULATED-TO-OBSERVED RATIOS

Most cross section ratio measurements in ISNF have been carried out relative

to either the 235U or the 239Pu fission reactions. Results of these measurements have been converted to reaction cross sections based on neutron fl uence

transfer from the 252Cf Fission Neutron Irradiation Facility employing the

239 Pu(n,f) reaction (Section 4. a). Cross sections for seven fissionable

isotopes and some significant capture reactions important for dosimetry and

nuclear technology are presented in Table 1-12.

Threshold reactions show discrepancies with calculated values similar to those

observed with fission spectra (cf. Tables X-10(B5) and Table X-17(B5) ), a

result consistent with the fact that the ISNF spectrum is dominated by the

fission spectrum in the MeV range. This similarity may be seen by comparing

truncated cross sections of threshold detectors in Table X- 1 3 ( B5 ) and

Table 1-5. The standard deviation of the twelve C/E ratios in Table 1-12,

excluding 232Th, is 0.028 about an average departure from unity of 1.03b.

99
Observed spectral indexes for ISNF are listed in column 2 of Table 1-13. In

column 3, they are given relative to the 235U fission spectrum. The 235U

fission spectrum values are from Table X-18, Part IB. The C/E ratios for the

spectral indexes are shown in Table 1-14. The assigned uncertainties

represent all error sources except the energy-dependent reaction cross

section. Errors associated with the physical and nuclear parameters of ISNF

are set out separately in the last two columns of the Table.

For the common fissile materials 239Pu and 235U, calculation and measurement

of the fission cross section ratio agree to within 1.5% indicating consistency

of the ENDF/B-V file between 1 keV and 3 MeV, the approximate energy response

range in ISNF for these two reactions. Poorer agreement for this ratio has

been noted for fission spectrum neutrons in Section 5.c, Part IB.

The C/E discrepancy of 5% for Syg/ygj the foremost spectral index in fastneutron technology, is similar to that observed for fission spectra

(C/E = 0.953 ± 0.024 in ISNF vs. 0.946 ± 0.015 in and 0.919 ± 0.022 in

x 25 ), even though the value of the index in ISNF is very different.

100
LIST OF TABLES

See

Secti on

Tabl e 1-1 Fourteen-Group Specification of the ISNF Central Spectrum 4

Tabl e 1-2 Spectrum Variation Corresponding to ISNF Parameter

Uncertai nties 4.b . 1

Tabl e 1-3 235U Fission Spectrum Errors from Part IA Compared to

Difference Between NBS and ENDF/B-V Evaluated Fission

Spectra 4.b.3

Tabl e 1-4 Assigned ISNF Spectrum and Errors 4 .b .3

Tabl e 1-5 Integral Detector Response Parameters 5 . a . 1

Tabl e 1-6 Cross Section Changes Associated with Individual ISNF

Parameter Uncertainties

5.a.l

Tabl e 1-7 Cross Section Changes Associated with 235U Fission

Spectrum Uncertainty

5.a.l

Tabl e 1-8 Total Propagated Error for Calculated ISNF Cross

Sections (1 S.D.)

5 . a .1

Tabl e 1-9 Calculated Spectral Indexes 5. a.

2

Table 1-10 Calculated Spectral Indexes —Truncated 5. a.

2

Tabl e 1-11 Calculated Spectral Index Errors Associated With

ISNF Spectrum Uncertainties (1 S.D.)

5. a.

2

Tabl e 1-12 Observed Integral Cross Sections 5.c.

Table 1-13 Observed Spectral Indexes 5.c.

Tabl e 1-14 Cal cul ated-to-Observed Ratios of Spectral Indexes 5 .c

.

101
Table 1-1. FOURTEEN-GROUP SPECIFICATION OF ISNF CENTRAL SPECTRUM

Designation: ISNF ( 5

) -

1 -LI

Lower

Energy

Boundary

Def i ned^ a ) ISNF

Spectrum

Ratio of Supplementary Calculations^)

to Reference Spectrum

Di screte

170 Group

ENDF/B-IV

(0RNL)

Ordi nates

40 Group

ENDF/B-III

(NBS)

Monte Carlo

ENDF/B-IV

(LASL)

Fission Spectrum Region

6.065 MeV 0.007374 1.003 0.996 1.065 ± .05

3.679 0.0362 1.002 1.023 0.971

Graphite Return Region

2.231 0.0948 1.011 1.035 0.990

1.353 0.1406 1.005 0.988 0.986

0.498 0.2574 1.002 0.993 0.985

0.1832 0.1800 1.000 0.997 1.006

10 Boron Absorption Region

111 keV 0.0632 0.999 0.992 1.002

67.38 0.0492 0.996 1.002 1.027

24.79 0.0705 0.991 1.005 1.015

9.119 0.0473 0.989 1.002 1.040

3.355 0.0287 0.983 0.992 1.004

1.234 0.01573 0.986 1.006 1.047

0.4540 0.00677 0.975 1.020 1.005 ± .05

0.414 eV 0.002178 1.001 1.076 1.113 ± .05

(^Results of Los Alamos 150-group ONEDANT calculation with ENDF/B-V cross sections (see Section 4.b.l). Spectrum in fine-group structure will be

furnished upon request.

(^The reference spectrum for these calculations is the 240-group LANL

spectrum described in section 4.b.l.

102
TABLE

1-2.

SPECTRUM

VARIATION

CORRESPONDING

TO

ISNF

PARAMETER

UNCERTAINTIES

5-S CD

a> as lo

£_ +-> +

13 •r- II 03 CO LO *=3" CM Xt* CO o CM CO xl" LO LO LO

+-> c JC 03 03 03 03 03 03 03 03 o O o o o o 03

(/> •»— Q. o 03 03 03 03 03 03 03 03 o o o o o o 03 • 1— TO t) o C- +-> O o o o O o o o t-H r—H t—H t—H t—H r—H o

Cl CCS Cl

v.<0 c CU 0O LO r-H xl- CO co 03 o CO LO CM 03 CO r— r-H o +-> CL oo CO 03 03 03 03 03 o o o O CO 03 03 t—H JD +-> --- 03 03 03 03 03 03 03 o o o o 03 03 03 o L_ fO CM

1X5

O CL o o O o o o o r-H t-H r-H t—H o o O r-H o OO E

a; CO r-H CO CO CO CO CO CO CO 00 o CO CO co o co o 13 1 03 03 03 03 03 03 03 03 03 o o CD o t—H o *1 03 03 03 03 03 03 03 03 03 o o o o CC o D -a ii O X3 o o O o O o o o o t—H t—H t—H t-H r—H t—H oo qc c.<3

s-s o E L_ t-H as + 00 03 o o r-H CO 03 o t—H r—H o 03 co co Xf

£Z +j

n 03 03 o o o 03 03 o o CD o 03 03 03 o •r— 4-> ^ 03 03 o o o 03 03 o o C o 03 03 03 o E rC o 3 u C> o o rH r-H r-H o o r-H r-H t—H t—H o o o r-H i

C/3 +-> c Q. ^ ^ <o

CM

-£= 1 O) -l-> 11 4-> I CO 03 03 CO co r-^ LO CO CO t—H 03 LO xj- CO xf

•r- SZ Cl o o o o o o o o o o 03 03 03 03 03 o

.c to i to o o o o o o o c o 03 03 03 03 03 o Cl CD as Qto s: as \ r-H rH t—H

r-H r-H T—

H

r—1

r-H t-H

O o o CO r—H

r-H

CCS Q_ ^<o C C\J o 4* LO LO LO LO LO H CO o co CO 03 t—H t—H CM CM CM CM CO CO

+-> ^ xj- CO CO co 03 03 03 o o O o o CD o 03

CO CL

O 03 03 03 03 03 03 03 o o CD CD c CO o 03 o £_ to

r_H

O +->

<S) Cl

OO o o o o o r-H t—H t—H t—H t—H t-H t—H o

-Oc£ lo

CD co CO CM

r

cj r-H CO LO CM CM o Xl" co oo CM CO

CLC CM CO LO CO o 03 CO o o Xf CO

d cdO o r-H CM Xl" CO CO LO Xj“ 03 CO o O 3

c_ ,

o o o o o o o o t—H CM r-H c O o

CD Ll_ o o o o o o o o o O o o O o

a> > > u

>5 > CD CD r-.

C

CD s: t—H as

£_ 03 TO | as c- -o Xl" Xt* LO CO r—H 03 LO , 2 <d c r-H LD CO CO CM LO CO r^. CO u. O C D Xl" CM CO » LU 1 CO xj- CO CM CO o o • • • • r—

CQ o O r-H CO 03 xl- t—H CO CO t—H CM CO CO CO CM co i—H CO 03 4—>

r—H t—H X^ o

i—

103

Factor

listed

is

ratio

of

the

group

fluence

calculated

with

ISNF

parameter

changed

as

indicated

to

the

group

fluence

with

the

parameter

unchanged.

The

reference

spectrum

for

these

studies

is

from

the

40-group
TABLE 1-3. 235 U FISSION SPECTRUM ERRORS FROM PART IA COMPARED TO DIFFERENCE

BETWEEN NBS AND ENDF/B-V EVALUATED FISSION SPECTRA

Lower

Energy

Bound

NBS

Spectrum

Eval uated

*Error

(la)

ENDF/B-V

NBS Eval

.

0.0 0.054 ± 16% 0.992

0.25 0.197 ± 4.1% 0.992

0.8 0.229 ± 3.0% 1.00

1.5 0.195 ± 3.1% 1.026

2.3 0.192 ± 2.0% 1.047

3.7 0.127 ± 4.8% 1.039

8 0.0056 ± 5.3% 1.089

Error based on departures of measured data sets — see Section 4.b.3 and

Table X-5 in Part IA.

104
TABLE 1-4. ASSIGNED ISNF SPECTRUM AND ERRORS

ERRORS

Lower Physical

Energy Spectrum and Nuclear Fission^

Boundary Parameters Spectrum

0.414 eV 0.00218 ± 6.3% ± 0.6%

0.454 keV 0.00677 ± 3.7 ± 0.6%

1.234 0.01573 ± 2.5

%

± 0.8%

3.355 0.0287 ± 2.7% ± 0.9%

9.12 0.0473 ± 1.4% ± 1.0%

24.8 0.0705 ± 1.0% ± 1.2%

67.4 0.0492 ± 0.7% ± 1.5%

111 0.0632 ± 0.5% ± 1.8%

183 0.1800 ± 0.41% ± 1.5%

498 0.2574 ± 0.6% ± 0.4%

1.353 MeV 0.1406 ± 0.6% ± 1.6%

2.231 0.0948 ± 1.4% ± 2.4%

3.679 0.0362 ± 1.4% ± 2.1%

6.065 to 17 0.00737 ± 1.4% ± 1.4%

Quadrature sum of errors listed in Table 1-2.

(b)

Error taken as one half of ISNF spectrum difference correspondi ng to

changing the source spectrum from the NBS evaluated 235U fission spectrum

to that of ENDF/B-V.

105
TABLE 1-5. INTEGRAL DETECTOR RESPONSE PARAMETERS

Spectrum : Intermediate-Energy Standard

Neutron Field (ISNF): Thick Shell

Cross Sections : ENDF/B-V Dosimetry File [EN79a]

Designation : I SNF ( 5

) -1-L1

Entry Date : June, 1983

Revi sed:

Detector

Reaction

Cross Section^ 3 )

CT 0 E p)

Spectrum^

)

F

r action

*(> e 95 )

Median^ 0 ) Response

Energy

E

p (p=0.5)

Response Ranged

Ep=0.4eV E p (p=0.95) E

p (p=0.95) Ep (p=0.05)

(barns) (barns) (MeV) (MeV) (MeV)

Spectrum Check

Constant , , 1.000 1.000 0.950 1.59 8.3 E-3 3.49

recip. vel . , (1 /v) 0.348 0.332 0.998 2.47 E-2 4.9 E-4 1.6

Fissionable Mat ' 1 s 239 Pu(n,f

)

1.821 1.750 0.989 0.58 1.5 E-3 3.5

235U(n,f

)

1.602 1.531 0.994 0.26 9.0 E-4 3.1

233U(n,f) 2.373 2.275 0.991 0.33 1.2 E-3 3.1

238U(n,f

)

0.1416 0.507 0.266 2.33 1.4 5.9

237 Np( n,f

)

0.803 1.464 0.521 1.54 0.53 4.5

232Th(n,f) 3.37 E-2 0.1221 0.2617 2.4 1.4 6.6

240 Pu(n,f

)

0.824 1.338 0.585 1.52 0.39 4.4

241 Pu(n ,f

)

2.102 2.010 0.994 0.26 9.4 E-4 3.0

238 U(n, y) 0.2215 0.2107 0.999 0.023 3.4 E-4 1.3

232Th(n , y) 0.2517 0.2394 0.999 0.031 3.2 E-4 1.5

Capture Reactions

23 Na 1.886 E-3 1.815 E-3 0.987 2.9 E-3 1.7 E-3 0.71

45 Sc 2.707 E-2 2.631 E-2 0.977 1.75 E-2 3.0 E-3 0.88

58 Fe 7.06 E-3 6.717 E-3 0.999 1.03 E-2 3.4 E-4 1.3

59 Co 4.09 E-2 3.883 E-2 1.000 1.42 E-4 1.23 E-4 0.94

63Cu 5.13 E-2 4.89 E-2 0.997 5.3 E-3 5.8 E-4 1.06

115 In 0.2838 0.2702 0.998 5.2 E-2 4.7 E-4 1.8

197Au 0.3924 0.3733 0.999 6.5 E-3 3.3 E-4 0.97

106
TABLE 1-5. INTEGRAL DETECTOR RESPONSE PARAMETERS (Continued)

Spectrum : Intermediate-Energy Standard

Neutron Field (ISNF): Thick Shell

Cross Sections: ENDF/B-V Dosimetry File [EN79a]

Designation : ISNF( 5) -1-L1

Entry Date : June, 1983

Revi sed

:

Detector

Reaction

Cross Section^ 3 ) °(> E

p

) Spectrum^)

Fraction

*(> E95)

Median^)

Response

Energy

Ep(p=0.5)

Response Ranged

Ep=0.4eV E p (p=0.95) E p (p=0.95) E p (p=0.05)

(barns) (barns) (MeV) (MeV) (MeV)

Helium Production

10 B(n,a) 1.667 1.587 0.998 2.03 E-2 4.7 E-4 1.8

6 Li (n,a) 0.7887 0.7535 0.995 0.205 8.4 E-4 3.3

Threshold Reactions

115 In(n,n ’

)

8.66 E-2 0.2283 0.360 2.28 1.0 5.3

47 Ti (n,p) 8.62 E-3 3.816 E-2 0.2145 3.33 1.7 7.0

32S(n,p) 2.552 E-2 0.1707 0.1403 3.69 2.2 7.0

58 Ni (n,p) 3.80 E-2 0.1954 0.1845 3.73 1.9 7.2

54 Fe(n,p) 2.845 E-2 0.1840 0.1469 3.87 2.2 7.3

46 Ti(n,p) 3.37 E-3 7.50 E-2 4.27 E-2 5.6 3.7 9.2

27 At(n,p) 1.293 E-3 2.333 E-2 5.27 E-2 5.6 3.4 9.2

56 Fe(n,p) 2.963 E-4 2.353 E-2 1.196 E-2 7.3 5.4 11

63 Cu(n,a) 1.608 E-4 7.13 E-3 2.143 E-2 7.3 4.7 11 27 At(n, a) 2.016 E-4 3.51 E-2 5.46 E-3 8.4 6.5 12

48 Ti ( n ,p) 7.95 E-5 8.93 E-3 8.46 E-3 8.1 5.9 12

Additions^

6 0 Ni ( n , p

)

7.53 E-4 3.79 E-2 1.890 E-2 6.9 4.8 10

55 Mn(n ,2n) 5.64 E-5 3.54 E-l 1.513 E-4 13 11 16

103 Rh(n,n'

)

0.390 0.689 0.538 1.76 0.49 4.9

9 3 Nb ( n , n '

)

0.0810 0.1793 0.429 2.18 0.79 4.9

dpa 500 667 0.712 1.8 0.19 5.3

107
Footnotes for TABLE I-5(B5)

^The value given in column 2 is the ful 1 -spectrum-averaged cross section

above a cadmium cut-off of 0.4 eV. The truncated cross section in column 3

is for a truncation energy (column 6) above which 95% of the detector

response occurs. A spectrum-averaged cross section truncated at energy E^

is given by

ff(> E

p

) = / a(EH(E)dE / *(E)dE

(b) The fraction of the spectrum above E 95

e E n (p = 0.95): ip(> E qc; ) =

9 5

'

/

ip( E ) dE . The ful 1 -spectrum-averaged cross section o{> 0.4 eV) is equal

•95

to a(> Eg 5

) • ip(> E q 0/0.95. 95 '

( c

)

v The fractions p = 0.95, 0.5, and 0.05 define energies above which 95%, 50%

(median), and 5% of the detector response occurs, respectively. E^ is

defined by the relation

/

a( E) E )dE = p

• [ a(> 0.4 eV)]

where E (p = 1) = 0.4 eV, and E(p = 0) = 20 MeV; and / ^(E)dE = 1.

p 0.4 eV

fdl

18 MeV

^ ^Normalization is /

a, , E\|>(E)dE = 1.

0.4 eV

17

( 6

)

Cross sections not taken from the ENDF/BV Dosimetry File:

- Nb(n,n‘) is from the IRDF Dosimetry File [Cu80a].

- Rh(n,n') is from data reported in Ref. [Pa80a],

- dpa, the atom displacement cross section, is from ASTM Standard Practice

E693-79 [An83a]

.

108
TABLE

1-6.

CROSS

SECTION

CHANGES

ASSOCIATED

WITH

INDIVIDUAL

ISNF

PARAMETER

UNCERTAINTIES

CMI

109

Parameter

uncertainties

are

given

in

column

headings

of

Table
TABLE 1-7. CROSS SECTION CHANGES ASSOCIATED WITH 238 U FISSION SPECTRUM UNCERTAINTY

Reaction E 50

Cross Section

ISNF (a ) F

i

ssi on

(b)

1JI,r Spectrum

Change

ISNF

Response

Above 0

ISNF

Fraction

.25 MeV

Fi ssi on Spectrum

Fiss. Spec.

Fi ssion (MeV)

239 Pu(

n

,f ) 0.6 +0.1% +0.1% <0.1% 0.64 0.95

235 U( n ,f ) 0.26 -0.2% 0.0% -0.2% 0.51 0.93

237 Np(n,f) 1.5 +1.0% +0.9% <0.2% 0.99 1.00

238 U(

n

,f ) 2.3 +1.7% +1.8% <0.2% 1.00 1.009

Non-Threshold (keV)

6 Li ( n , a) 205 -0.7 (-1.1%) 0.39 0.83

10 B( n , a) 20 -0.6 (-0.9%) 0.19 0.80

238 U( n , y) 23 -0.5% (-1.3%) 0.25 0.86

197 Au(n, y) 6.5 -0.6% (-1.7%) 0.14 0.79

23 Na(n,Y) 2.9 -0.5% (-1.2%) 0.097 0.75

Threshold (MeV)

115 In(n,n'

)

2.3 +1.6% +1.7% <0.2% 1.00 1.00

58 Ni ( n ,p) 3.7 +1.8% +2.0% -0.2% 1.00 1.00

*t6 Ti(n,p) 5.6 +1.4% +1.6% -0.2% 1.00 1.00

48 Ti (n,p) 8.1 +1.6% +1.7% <0.2% 1.00 1.00

( d )one half of departure from unity in percent of the cross section obtained

with an ISNF calculation using the ENDF/B-V fission spectrum divided by the

cross section obtained with a calculation using the NBS-eval uated fission

spectrum.

(k)one half of departure from unity of the fission spectrum cross section

obtained using the ENDF/B-V fission spectrum evaluation divided by the cross section obtained with the NBS evaluation.

110
TABLE 1-8. TOTAL PROPAGATED ERROR FOR CALCULATED ISNF CROSS SECTIONS (1 S.D.)

Reaction

Median

Response

Energy

C E

5 0 )

Total Error

Component From

ISNF Parameters

(Table 1-6)

Total Propagated Error

Cross Section^)

Absolute^ 3 ' Relative to

Cross Section Fission Spectrum

Fission (MeV)

239p u 0.6 ± 0.3% ± 0.3% ± 0.3%

235 Us 2 3 3y 0.3 ± 0.6% ± 0.6% ± 0.6%

237 Np, 240 Pu 1.5 ± 0.6% ± 1.2% ± 0.6%

238 U, 232Jh 2.3 ± 1.0% ± 2.0% ± 1.0%

Non-Threshol

d

(keV)

6 Li (n, a) 205 ± 0.6% ± 0.9% -

10 B(n,a) 20 ± 1.4% ± 1.5% - 238 U( n , y) 23 ± 1.4% ± 1.5% - 197Au(n,Y) 6.5 ± 1.9% ± 2.0% - 23 Na(n, y) 2.9 ± 1.8% ± 1.9% -

Threshol

d

(MeV)

115 In(n,n'

)

2.3 ± 0.9% ± 1.8% ± 0.9%

58 Ni (n,p) 3.7 ± 1.4% ± 2.3

%

± 1.4%

46Ti(n,p) 5.6

48Ti (n,p) 8.1 ± 1.4% ± 2.1% ± 1.4%

(^Quadrature sum of errors in column 3 and (Tabe 1-7, column 3).

(^Quadrature sum of error in column 3 and (Table 1-7, column 5).

Ill
TABLE 1-9. CALCULATED SPECTRAL INDEXES

Spectrum:

Spectral

Intermedi ate-Energy Standard Designation : ISNF (5) -1-L1

Neutron Field (ISNF): Thick Shell

Entry Date : June, 1983

Index : Ful 1 -Spectrum Cross Sections from

Table 1-5 (ENDF/B-V Dosimetry File) Revised:

^a/B^calc. “ t^c/^calc.

112
TABLE 1-10. CALCULATED SPECTRAL INDEXES — TRUNCATED

Spectrum : Intermedi ate-Energy Standard

Neutron Field (ISNF): Thick Shell

Spectral Index : Truncated Cross Sections from

Table 1-5 (ENDF/B-V Dosimetry File)

Designation : ISNF ( 5

) -1-L1

Entry Date: June, 1983

Revi sed

:

113
TABLE 1-11. CALCULATED SPECTRAL INDEX ERRORS ASSOCIATED WITH ISNF

SPECTRUM UNCERTAINTIES (1 S.D.)

Propagated Error^ a ^

Spectral Index

Without

Normalization to

Fission Spectrum

With

Normalization to

Fission Spectrum

Fission

235 U/3

( 3= 239 Pu, 233U) < ±0.5% < ±0.5%

235 U/

3

( 3= 238 U, 232Th) ± 2.6

%

± 1.4%

235U/ 237 Np ± 1.4% ± 1.0%

2 3

7

Np/ 238U ± 0.9% ± 0.7%

Non-Threshol

d

6 Li (n,a)/ 235U(n,f) ± 0.5% -

10 B(n, a)/ 235U(n,f

)

± 0.9% -

238 U(n, y)/3

( 3= 235 U, 239Pu , 233U fission) ± (0.9 to 1.3)% -

197 Au( n, y)/

3

( 3= 235 U, 239Pu , 233U fission) ± (1.4 to 1.9)% -

238 U(n,-y)/ 238U(n,f) ± 2.9

%

-

Threshol

d

115 In(n,n ' )/ 238U(n,f

)

± 0.3% ± 0.2%

58 Ni (n,p)/ 238U(n,f

)

± 0.5% ± 0.5%

58 Ni (n,p)/ 239Pu(n,f

)

± 2.1% ± 1.3%

48Ti (n,p)/ 238U(n,f) ± 0.7% ± 0.7%

(^Algebraic sum of correspondi ng errors in Table 1-6 and 1-7 added in

quadrature —see Section 5. a. 2.

114
TABLE 1-12. OBSERVED INTEGRAL CROSS SECTIONS

Spectrum : Intermediate-Energy Standard Desi gnation : I SNF ( 5

) -1 -LI

Neutron Field (ISNF): Thick Shell

Entry Date : June, 1983

Revi sed : July, 1985

Median

Reaction Response Cross Section' 3 ' Calculated-'' 0 '' Energy, Value to-Observed References

E 50 (MeV) ( xio 27 cm 2

)

Fi ssion

2 3 5 u 0.26 1606 + 2.2% 0.998 + 2.3% [Gr83a], [Wa81a]

2 3 3y 0.33 2424 + 2.7% 0.980 + 2.8% [Gr83a]

238u 2.33 149.0 + 2.4% 0.950 + 3.1% [Gr83a], [Wa81a]

2 3 7 Np 1.54 829 + 2.7% 0.969 + 3.0% [Gr83a], [Wa81a]

232 Th 2.4 38.4 + 3.2% 0.878 + 3.8% [Gr83a]

240p u 1.52 824 + 2.8% 1.000 + 3.0% [Gr83a], [Mc82a]

241 Pu 0.26 2152 + 5.0% 0.977 + 5.0% [Gr83a]

Non-Threshol

d

10 B(n,He) 0.020 1831 + 3.3% 0.910 + 3.6% [Gr85a] , [0184a]

6 Li ( n,He) 0.205 831 + 3.0% 0.949 + 3.1% [Gr85a] , [0184a]

238 U(n, y) 0.023 227 + 2.6 % 0.976 + 3.0% [Gi 79a]

232Th(n,y) 0.031 265 + 3.6% 0.950 + 3.9% [Gi 79a]

197Au( n, y) 0.0065 398 + 3.4% 0.986 + 3.7% [Fa79a]

Threshol d

115 In(n,n'

)

2.3 91.8 + 3.0% 0.943 + 3.5% [Fa79a]

(^Derived from cross section ratios of Table 1-13 and fluence transfer from

252Cf Fission Neutron Irradiation Facility. The fluence transfer error is ± 2.1%—see Section 4. a.

(^Calcul ated cross section from Table 1-5 (ENDF/B-V) with error from column

of Table 1-8 combined in quadrature with cross section error in column

115
TABLE 1-13. OBSERVED SPECTRAL INDEXES

Spectrum: Intermedi ate-Energy Standard Designation: ISNF ( 5) -1-L1

Neutron Field (ISNF): Thick Shell

Entry Date : June, 1983

Revised: July, 1985

^a/B^obs ~ [^o/°B-lobs

Reaction

Spectral

(

a ) Index

1 s o/bJisnf

Spectral Index^ b ^ Relative to 235 U Fission Spectrum

l S a/3hsNF / [ S a/ 3^235

References

Fi ssi on 235y/239pu 0.866 ± 1.0% 1.312 ± 1.4% [Wa81a]

23 3y/ 23 5y 1.509 ± 1.6% 0.930 ± 2.1% [Gr83a]

23 8 y/ 23 5y 0.0928± 1.1% 0.357 ± 1.2% [Wa81a]

237 Np/ 2 3 5 U 0.516 ± 1.6% 0.456 ± 1.8% [WaSla]

240pu / 235y 0.513 ± 1.8% 0.463 ± 2.1% [Mc82a]

241 Pu/ 235 U 1.340 ± 4.6% - [Gr83a]

Non-Threshol

d

10 B(n,He)/ 235 U(n,f) 1.141 ± 2.7% - [Gr85a]

6

Li ( n ,He)/ 235 U(n ,f

)

0.518 ± 2.3

%

- [Gr85a]

238 U(n,y)/ 235 U(n,f) 0.1412± 1.7% - [Gi 79a]

232Th(n,y)/ 235 U(n,f) 0.165 ± 3 %

- [Gi 79a]

197 Au(n,y)/ 235 U(n,f) 0.248 ± 2.9

%

- [Fa79a]

Threshold

1 15 In( n ,n

1 )/

238 U( n ,f

)

0.616 ± 2.1% 1.011 ± 2.9% [Fa79a]

(^Observed values obtained with activation detectors or an NBS double fission

chamber.

(b)spectral indexes for the 235 U fission spectrum are from Table X-18,

Part IB.

116
TABLE 1-14 CALCULATED-TO-OBSERVED RATIOS OF SPECTRAL INDEXES

Spectrum: Intermedi ate-Energy Standard

Neutron Field (ISNF): Thick Shell

Desi gnation : ISNF( 5) -1-L1

Entry Date : June 1983

Revised: July, 1985

Cal cul ated-to-Observed

Reacti ons

( a ) Without (b)with

Normalization to Normalization to

23 5y 23 5y

Fission Spectrum Fission Spectrum

Spectrum-Associ ated

Error in

Calculated Index

(Table 1-11)

Fi ssi on w/o norm w/norm

235u/239pu 1.015 + 1.1% 0.969 + 1.5% < + 0.5% < ± 0.5%

23 3yy 2 3 5

y

0.981 + 1.7% 1.029 + 2.2% < + 0.5% < ± 0.5%

23 8y/ 23 5y 0.953 + 2.5% 1.039 + 1.9% + 2.3% ** +i pH •

2 3 7 Np/ 2 3 5 U 0.971 + 2.1% 1.027 + 2.1% + 1.4% • o

f +i 240pu /

235 U 1.002 + 2.3% 1.035 + 2.3% + 1.4% 1+ »

—

• o3^ 241p u /

235 u 0.979 + 4.6% - < + 0.5%

Non-Threshold

10 B(n,a)/ 235 U(n,f) 0.912 + 2.9% - ± 0.9% -

6

Li (n, a)/ 235 U(n,f) 0.950 + 2.4% - ± 0.5% - 238 U(n, y)/ 235 U(n,f) 0.979 + 2.0% - ± 1.0% - 232Th(n,y)/ 235 U(n,f) 0.952 + 3.2% - ± 1.0% - 197 Au(n, y)/ 235 U(n,f) 0.988 + 3.2% - ± 1.4% -

Threshold

1 15 In(n ,n

1 )/ 238U( n ,f

)

0.994 + 2.1% 1.029 + 3.7% ± 0.3% ± 0.2%

(^Calculated spectral index from Table 1-9; observed value from Table l-i , column 2. Includes ISNF spectrum error from column 4.

(^Calculated spectral indexes from Table 1-9 (ISNF) and Table X-14(B5),

Part IB; observed spectral indexes from Table 1-13, column 3. IncHdfs

spectrum error from column 5.

117
CHANNEL

FOR

BORAL

CURTAIN

)*: >‘ Q’- :~ P. \/ :\

. -V. *.: \v VV w Vv

: ? &&-<?:&

* , -= ‘vc

O; -V,’

•:fr& o .\\ ?\

' o x ’« b , p

sis: ss ' v

~v~^

r\ QC

CD

CD

CD

118

Figure

1-1

.

Cavity

arrangement

in

the

NBS

reactor

thermal

column.
CAVITY

(30

cm

dig.)

119

Figure

1-2.

Intermediate-Energy

Standard

Neutron

Field

(ISNF)

assembly

at

the

center

of

the

NBS

Reactor

thermal

col

umn

.
BORON

ALUMINUM

FISSION

CARBON

SHELL

CAP

8

RING

SOURCE

*

\

/

REGION

\

E

o

120

Figure

1-3.

Geometry

and

material

densities

for

one-dimensional

transport

calculation

of

central

ISNF

neutron

spectrum.
.50

oinoinoinoinoino in -c* 0"> o-> cm cm — — o

CM

o

in

CM

o

o

m

>

UJ CM C

LU

Io

in

CM

CM

IO

in

CM

fn

IO

in o in

** oo o in cn cm

(3)<#> 3

o in cm — o in — o

o

121

Figure

1-4.

Central

flux

lethargy

spectrum

for

ISNF

in

240

energy

groups.
122

NEUTRON

ENERGY

E,

MeV
o

o

123

•i^ure

1-6.

Comparison

of

single-scatter

and

discrete-ordinates

calculation

of

aluminum

resonance

scattering

in

the

boron

shell.



REFERENCES

[Ac81a] Acquah, I., Glumac, B., Remec, I., and Najzer, M., "Spectral Indices

of Some Threshold Reactions Measured in Uranium-235 Fission Spectrum,"

Proc. Advisory Group Meeting on Nuclear Data for Radiation Damage

Assessment and Related Safety Aspects, IAEA-TECDOC-263 , Vienna (Oct.

1981).

[Ad77a] Adamov, V. M., Alexandrov, B. M., Alkhazov, I. D., Drapchinsky, L. V.,

Kovalenko, S. S., Kostochkin, 0. I., Kudriavzev, G. Yu., Malkin, L.Z.,

Petrzhak, K. A., PI eskachevsky , L. A., Fomichev, A. V., and Shapakov,

V. I., "Absolute 235 U, 238 U, 237 Np Fast Neutron Fission Cross-Section

Measurements," Proc. Symposium on Neutron Standards and Applications,

March 1977, Washington, D.C., NBS Special Publication 493.

[A175a] Alberts, W. G., Bortfeldt, J., GClnther, E., Knauf, K., Matzke, M.,

Rassl , G., Siegel, V., and Walz, K. F., "Measurement of Cross Sections

for Threshold Reaction Induced by 252Cf Spontaneous Fission Neutrons,"

Proceedings Nuclear Cross Sections and Technology Conference, NBS

Special Publication 425, Vol . I, p. 273 (1975).

[A1 75b] Alberts, W. G., GCfnther, E., Matzke, M., and Rassl, G., "Measurements

of Integral Cross Sections in the 252Cf Fission Spectrum," Proc. of

First ASTM-EURATOM Symposium on Reactor Dosimetry, EUR 5667 e/f, Part

II, p. 131, Petten, (September 22-26, 1975).

[A182a] Alberts, W. G., Hollnagel, R., Knauf, K., Matzke, M., and Pebara, W.,

"Measurements with the Niobium Neutron Fluence Detector at the PTB,"

Proc. Fourth ASTM-EURATOM Symposium on Reactor Dosimetry, Washington,

D.C., NUREG/CP-0029 , Vol. 1, 433 (March 1982).

[An83a] 1983 Annual Book of ASTM Standards, Section 12, Vol. 12.02,

Nuclear (II), Solar, and Geothermal Energy, Standard Practice E693-79;

American Society for Testing and Materials, Philadelphia, PA (1983).

[Br70a] Bresesti , A., Bresesti , M., Rota, A., and Rydin, R., "Threshold

Reaction Excitation Functions Intercal i brated in a Pure Fission

Spectrum," Nucl . Sci . Eng. 40, 331 (1970).

[Br79a] Broadhead, B. L. and Wagschal , J., "The NBS Intermediate-Energy

Standard Neutron Field (ISNF) Revisited," Trans. Am. Nucl. Soc. 33

(November, 1979).

[Br79b] Broadhead, B., Private communication (February, 1979).

[Br80a] Broadhead, B. L. and Wagschal, J. J., "The ISNF: A New CSEWU Dosinetr,

Benchmark (Computational Problems and Their Solution)," Trans. An.

Nucl . Soc. 35, p. 466 (1980)

.

[Br81a] Brandon, W. E., Cogburn, C. 0., Culp, R. R., Sallee, W. W., anc Williams, J. G., "Wall Return of 252Cf Neutrons at the SEFOR Cali:>r<ition Center," Trans. Am. Nucl. Soc. 38, 579 (June, 1981 ).

125
[Br81b] Brandon, W. E., Cogburn, C. 0., Culp, R. R., Sallee, W. W., and

Williams, J. G., "Calibration of Neutron Dosimetry Techniques For Use

in Reactor Buildings," Proc. Eighth DOE Workshop on Personnel Neutron

Dosimetry, PNL-SA-9950, Louisville, Kentucky (June, 1981).

[Ca84a] Carlson, A. D., "Standard Cross-Section Data," Progress in Nuclear

Energy, 13, No. 2/3, 95-102 (1984).

[Cr56a] Cranberg, L., Frye, G., Nereson, N., and Rosen, L., Phys. Rev. 103,

662 (1956); see also, Barnard, E., et al . , Nucl . Phys. 71, 228 (1965);

Watt, R., Phys. Rev. 87, 1037 (1952).

i

—

i

o c oo 0 Cu

1

i Cullen, D. E., Kucherov, N., and McLaughlin, P. M., "International

Reactor Dosimetry File-82," IAEA-NDS-41/R. See also Strohmaier, B.,

Tagesen, S., and Vonach, H., Physics Datem-Physi cs Data 13 (1980).

[Da78a] Davis, M. C., and Knoll, G. F., "Fission Cross Sections of 235U and

239 Pu Averaged Over 252Cf Neutron Spectrum," Ann. Nucl. Energy 5, 583

(1978).

[Ei 77a] Eisenhauer, C. M. and Grundl , J. A., "Neutron Transport Calculations

for the Intermedi ate-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).

[ EN79a] ENDF/B-V, Report BNL-17541 (ENDF-201), edited by R. Kinsey, available

from the National Nuclear Data Center, Brookhaven National Laboratory,

Upton, NY (July 1979); See also Magurno, B. A., "ENDF/B-IV Dosimetry

File," Report BNL-NCS-50446 (1975).

[Fa72a] Fabry, A., "Evaluation of Microscopic Integral Cross Section Averaged

in a 235 U Fission Neutron Spectrum," Report Big 465, CEN/SCK, Mol

(1972).

[ Fa72b] Fabry, A. and Jenkins, J. D., "Wall Return Neutron Fluxes for High and

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

J_5, 2 ( 1972).

[Fa74a] Fabry, A. and Czock, K., "Measurements of Integral Cross Section

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