ASTM E844-97
(Guide)Standard Guide for Sensor Set Design and Irradiation for Reactor Surveillance, E 706(IIC)
Standard Guide for Sensor Set Design and Irradiation for Reactor Surveillance, E 706(IIC)
SCOPE
1.1 This guide covers the selection, design, irradiation, post-irradiation handling, and quality control of neutron dosimeters (sensors), thermal neutron shields, and capsules for reactor surveillance neutron dosimetry.
1.2 The values stated in inch-pound units are to be regarded as the standard. The values given in parentheses are for information only.
1.3 This standard does not purport to address all of the safety problems, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.
General Information
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Standards Content (Sample)
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Designation: E 844 – 97
Standard Guide for
Sensor Set Design and Irradiation for Reactor Surveillance,
E 706(IIC)
This standard is issued under the fixed designation E 844; the number immediately following the designation indicates the year of
original adoption or, in the case of revision, the year of last revision. A number in parentheses indicates the year of last reapproval. A
superscript epsilon (e) indicates an editorial change since the last revision or reapproval.
1. Scope 3. Terminology
1.1 This guide covers the selection, design, irradiation, 3.1 Definitions:
post-irradiation handling, and quality control of neutron do- 3.1.1 neutron dosimeter, sensor, monitor—a substance irra-
simeters (sensors), thermal neutron shields, and capsules for diated in a neutron environment for the determination of
reactor surveillance neutron dosimetry. neutron fluence rate, fluence, or spectrum, for example: radio-
1.2 The values stated in inch-pound units are to be regarded metric monitor (RM), solid state track recorder (SSTR), helium
as the standard. The values given in parentheses are for accumulation fluence monitor (HAFM), damage monitor
information only. (DM), temperature monitor (TM).
1.3 This standard does not purport to address all of the 3.1.2 thermal neutron shield—a substance (that is, cad-
safety problems, if any, associated with its use. It is the mium, boron, gadolinium) that filters or absorbs thermal
responsibility of the user of this standard to establish appro- neutrons.
priate safety and health practices and determine the applica- 3.2 For definitions or other terms used in this guide, refer to
bility of regulatory limitations prior to use. Terminology E 170.
2. Referenced Documents 4. Significance and Use
2.1 ASTM Standards: 4.1 In neutron dosimetry, a fission or non-fission dosimeter,
E 170 Terminology Relating to Radiation Measurements or combination of dosimeters, can be used for determining a
and Dosimetry fluence-rate, fluence, or neutron spectrum, or both, in nuclear
E 261 Practice for Determining Neutron Fluence Rate, Flu- reactors. Each dosimeter is sensitive to a specific energy range,
ence, and Spectra by Radioactivation Techniques and, if desired, increased accuracy in a flux-spectrum can be
E 854 Test Method for Application and Analysis of Solid achieved by the use of several dosimeters each covering
State Track Recorder (SSTR) Monitors for Reactor Sur- specific neutron energy ranges.
veillance, E 706(IIIB) 4.2 A wide variety of detector materials is used for various
E 910 Test Method for Application and Analysis of Helium purposes. Many of these substances overlap in the energy of
Accumulation Fluence Monitors for Reactor Vessel Sur- the neutrons which they will detect, but many different
veillance, E 706(IIIC) materials are used for a variety of reasons. These reasons
E 1005 Test Method for Application and Analysis of Radio- include available analysis equipment, different cross sections
metric Monitors for Reactor Vessel Surveillance, for different flux levels and spectra, preferred chemical or
E 706(IIIA) physical properties, and, in the case of radiometric dosimeters,
E 706(IIID) Analysis of Damage Monitors for Reactor Ves- varying requirements for different half-life isotopes, possible
sel Surveillance interfering activities, and chemical separation requirements.
E 706(IIIE) Analysis of Temperature Monitors for Reactor
3 5. Selection of Neutron Dosimeters and Thermal Neutron
Vessel Surveillance
Shields
E 706(IIE) Benchmark Testing of Reactor Vessel Dosim-
etry 5.1 Neutron Dosimeters:
5.1.1 The choice of dosimeter material depends largely on
the dosimetry technique employed, for example, radiometric
This guide is under the jurisdiction of ASTM Committee E-10 on Nuclear
monitors, helium accumulation monitors, track recorders, and
Technology and Applications and is the direct responsibility of Subcommittee
damage monitors. At the present time, there is a wide variety of
E10.05 on Nuclear Radiation Metrology.
detector materials used to perform neutron dosimetry measure-
Current edition approved June 10, 1997. Published May 1998. Originally
published as E 844 – 81. Last previous edition E 844 – 86 (1991). ments. These are generally in the form of foils, wires, powders,
Annual Book of ASTM Standards, Vol 12.02.
and salts. The use of alloys is valuable for certain applications
For standards that are in the draft stage and have not received an ASTM
designation, see Section 5 as well as Figures 1 and 2 of Matrix E 706.
Copyright © ASTM, 100 Barr Harbor Drive, West Conshohocken, PA 19428-2959, United States.
NOTICE: This standard has either been superceded and replaced by a new version or discontinued.
Contact ASTM International (www.astm.org) for the latest information.
E 844
such as (1) dilution of high cross-section elements, (2) prepa- there is little or no change in neutron spectral shape or fluence
ration of elements that are not readily available as foils or wires
rate with time.
in the pure state, and (3) preparation to permit analysis of more
5.1.9 Tables 1-3 present various dosimeter elements. Listed
than one dosimeter material.
are the element of interest, the nuclear reaction, and the
5.1.2 For neutron dosimeters, the reaction rates are usually
available forms. For the intermediate energy region, the ener-
deduced from the absolute gamma-ray radioanalysis (there
gies of the principal resonances are listed in order of increasing
exist exceptions, such as SSTRs, HAFMs, damage monitors).
energy. In the case of the fast neutron energy region, the 95 %
Therefore, the radiometric dosimeters selected must have
response ranges (an energy range that includes most of the
gamma-ray yields known with good accuracy (>98 %). The
response for each dosimeter is specified by giving the energies
half-life of the product nuclide must be long enough to allow
E below which 5 % of the activity is produced and E above
05 95
for time differences between the end of the irradiation and the
which 5 % of the activity is produced) for the U neutron
subsequent counting. Refer to Method E 1005 for nuclear
thermal fission spectrum are included.
decay and half-life parameters.
5.2 Thermal Neutron Shields:
5.1.3 The neutron dosimeters should be sized to permit
5.2.1 Shield materials are frequently used to eliminate
accurate analysis. The range of high efficiency counting
interference from thermal neutron reactions when resonance
equipment over which accurate measurements can be per-
and fast neutron reactions are being studied. Cadmium is
formed is restricted to several decades of activity levels (5 to 7
commonly used as a thermal neutron shield, generally 0.020 to
decades for radiometric and SSTR dosimeters, 8 decades for
0.050 in. (0.51 to 1.27 mm) thick. However, because elemental
HAFMs). Since flux levels at dosimeter locations can range
cadmium (m.p. = 320°C) will melt if placed within the vessel
over 2 or 3 decades in a given experiment and over 10 decades
of an operating water reactor, effective thermal neutron filters
between low power and high power experiments, the proper
must be chosen that will withstand high temperatures of
sizing of dosimeter materials is essential to assure accurate and
light-water reactors. High-temperature filters include cadmium
economical analysis.
oxide (or other cadmium compounds or mixtures), boron
5.1.4 The estimate of radiometric dosimeter activity levels
(enriched in the B isotope), and gadolinium. The thicknesses
at the time of counting include adjustments for the decay of the
of the shield material must be selected to account for burnout
product nuclide after irradiation as well as the rate of product
from high fluences.
nuclide buildup during irradiation. The applicable equation for
5.2.2 In reactors, feasible dosimeters to date whose response
such calculations is (in the absence of flux perturbations) as
range to neutron energies of 1 to 3 MeV includes the fission
follows:
238 237 232
monitors U, Np, and Th. These particular dosimeters
lt1 2lt2
A 5 N s¯fa~1 2 e !~e ! (1)
o
must be shielded from thermal neutrons to reduce fission
235 238
product production from trace quantities of U, Pu, and
where:
Pu and to suppress buildup of interfering fissionable nu-
A = expected disintegration rate (dps) for the prod-
238 238 237
clides, for example, Np and Pu in the Np dosimeter,
uct nuclide at the time of counting,
239 238 233 232
Pu in the U dosimeter, and Uinthe Th dosimeter.
N = number of target element atoms,
o
Thermal neutron shields are also necessary for epithermal
f = estimated flux density level,
−7
spectrum measurements in the 5 3 10 to 0.3-MeV energy
s¯ = spectral averaged cross section,
a = product of the nuclide fraction and (if appli- range. Also, nickel dosimeters used for the fast activation
58 58
cable) of the fission yield, reaction Ni(n,p) Co must be shielded from thermal neutrons
-lt1
1−e = buildup of the nuclide during the irradiation
in nuclear environments having thermal fluence rates above 3
12 −2 −1 58 58m
period, t , 3 10 n·cm ·s to prevent significant loss of Co and Co
-lt2
e = decay after irradiation to the time of counting,
t , and
l = decay constant for the product nuclide.
TABLE 1 Dosimeter Elements—Thermal Neutron Region
5.1.5 For SSTRs and HAFMs, the same type of information
Element of
Nuclear Reaction Available Forms
as for radiometric monitors (that is, total number of reactions)
Interest
10 7
is provided. The difference being that the end products (fission
B B(n,a) Li B, B C, B-Al, B-Nb
59 60
Co Co(n,g) Co Co, Co-Al, Co-Zr
tracks or helium) require no time-dependent corrections and
63 64
Cu Cu(n,g) Cu Cu, Cu-Al, Cu(NO )
3 2
are therefore particularly valuable for long-term irradiations.
197 198
Au Au(n,g) Au Au, Au-Al
115 116m
5.1.6 Fission detectors shall be chosen that have accurately
In In(n,g) In In, In-Al
58 59
Fe Fe(n,g) Fe Fe
known fission yields. Refer to Method E 1005.
54 55
Fe Fe(n,g) Fe Fe
5.1.7 In thermal reactors the correction for neutron self
6 3
Li Li(n,a) H LiF, Li-Al
55 56
shielding can be appreciable for dosimeters that have highly
Mn Mn(n,g) Mn alloys
58 59 56
Ni Ni(n,g) Ni(n,a) Fe Ni
absorbing resonances (see 6.1.1).
Pu Pu(n,f)FP PuO , alloys
5.1.8 Dosimeters that produce activation or fission products 45 46
Sc Sc(n,g) Sc Sc, Sc O
2 3
109 110m
(that are utilized for reaction rate determinations) with half- Ag Ag(n,g) Ag Ag, Ag-Al, AgNO
23 24
Na Na(n,g) Na NaCl, NaF, NaI
lives that are short compared to the irradiation duration should
181 182
Ta Ta(n,g) Ta Ta, Ta O
2 5
not be used. Generally, radionuclides with half-lives less than 235
U (enriched) U(n,f)FP U, U-Al, UO ,U O , alloys
2 3 8
three times the irradiation duration should be avoided unless
NOTICE: This standard has either been superceded and replaced by a new version or discontinued.
Contact ASTM International (www.astm.org) for the latest information.
E 844
TABLE 2 Dosimeter Elements—Intermediate Neutron Region
Energy of Principal
Resonance, eV Dosimetry Reactions Element of Interest Available Forms
(17)
A 6 3
Li(n,a) H Li LiF, Li-Al
A 10 7
B(n,a) Li B B, B C, B-Al, B-Nb
A 58 59 56
Ni(n,g) Ni(n,a) Fe Ni Ni
115 116m
1.457 In(n,g) In In In, In-Al
181 182
4.28 Ta(n,g) Ta Ta Ta, Ta O
2 5
197 198
4.906 Au(n,g) Au Au Au, Au-Al
109 110m
5.19 Ag(n,g) Ag Ag Ag, Ag-Al, AgNO
232 233
21.806 Th(n,g) Th Th Th, ThO , Th(NO )
2 3 4
B 235
U(n,f)FP U U, U-Al, UO ,U O , alloys
2 3 8
59 60
132 Co(n,g) Co Co Co, Co-Al, Co-Zr
58 59
1038 Fe(n,g) Fe Fe Fe
55 56
337.3 Mn(n,g) Mn Mn alloys
63 64
579 Cu(n,g) Cu Cu Cu, Cu-Al, Cu(NO )
3 2
0.2956243 Pu(n,f)FP Pu PuO , alloys
23 24
2810 Na(n,g) Na Na NaCl, NaF, NaI
45 46
3295 Sc(n,g) Sc Sc Sc, Sc O
2 3
54 55
7788 Fe(n,g) Fe Fe Fe
A
This reaction has no resonance that contributes in the intermediate energy region and the principle resonance has negative energy (i.e. the cross section is 1/v).
B
Many resonances contribute in the 1 – 100 eV region for this reaction.
TABLE 3 Dosimeter Elements—Fast Neutron Region
A,B
Energy Response Range (MeV)
Cross Section
Dosimetry Element of Available
Uncertainty
Low Median High
Reactions Interest Forms
A,C
(%)
E E E
05 50 95
Np(n,f)FP Np 0.684 1.96 5.61 9.33 Np O , alloys
2 3
103 103m
Rh(n,n8) Rh Rh 0.731 2.25 5.73 3.1 Rh
93 93m
Nb(n,n8) Nb Nb 0.951 2.57 5.79 3.06 Nb, Nb O
2 5
115 115m
In(n,n8) In In 1.12 2.55 5.86 2.17 In, In-Al
14 11
N(n,a) B N 1.75 3.39 5.86 — TiN, ZrN, NbN
U(n,f)FP U (depleted) 1.44 2.61 6.69 0.53 U, U-Al, UO ,U O , alloys
3 3 8
Th(n,f)FP Th 1.45 2.79 7.21 5.09 Th, ThO
9 6
Be(n,a) Li Be 1.59 2.83 5.26 — Be
47 47
Ti(n,p) Sc Ti 1.70 3.63 7.67 2.17 Ti
58 58
Ni(n,p) Co Ni 1.98 3.94 7.51 2.43 Ni, Ni-Al
54 54
Fe(n,p) Mn Fe 2.27 4.09 7.54 2.17 Fe
32 32
S(n,p) P S 2.28 3.94 7.33 4.0 CaSO ,Li SO
4 2 4
32 29
S(n,a) Si S 1.65 3.12 6.06 — Cu S, PbS
58 55
Ni(n,a) Fe Ni 2.74 5.16 8.72 — Ni, Ni-Al
46 46
Ti(n,p) Sc Ti 3.70 5.72 9.43 2.46 Ti
56 56 D
Fe(n,p) Mn Fe 5.45 7.27 11.3 2.33 Fe
56 53
Fe(n,a) Cr Fe 5.19 7.53 11.3 — Fe
63 60 E
Cu(n,a) Co Cu 4.53 6.99 11.0 2.85 Cu, Cu-Al
27 24
Al(n,a) Na Al 6.45 8.40 11.9 1.40 Al, Al O
2 3
48 48
Ti(n,p) Sc Ti 5.92 8.06 12.3 6.66 Ti
47 44
Ti(n,a) Ca Ti 2.80 5.10 9.12 — Ti
60 60 F E
Ni(n,p) Co Ni 4.72 6.82 10.8 10.3 Ni, Ni-Al
55 54 G
Mn(n,2n) Mn Mn 11.0 12.6 15.8 13.4 alloys
A 235
Energy response range was derived using the ENDF/B-VI U fission spectrum, reference (17), MT = 9228, MF = 5, MT = 18. The cross section and associated
covariance sources are identified in standard E 1018 and in Reference (18).
B 50
One half of the detector response occurs below an energy given by E ; 95 % of the detector response occurs below E and 5 % below E .
95 05
C
Uncertainty metric only reflects that component due to the knowledge of the cross section and is reported at the 1s level.
D
Low manganese content necessary.
E
Low cobalt content necessary.
F
This reaction does not appear in the ENDF/B-VI evaluation. The JENDL evaluation (19) was used.
G
Low iron content necessary.
by thermal neutron burnout (1). personnel safety and accurate nuclear environment character-
ization. During dosimeter fabrication, care must be taken in
6. Design of Neutron Dosimeters, Thermal Neutron
order to achieve desired neutron flux results, especially in the
Shields, and Capsules
case of thermal and resonance-region dosimeters. A number of
factors must be considered in the design of a dosimetry set for
6.1 Neutron Dosimeters—Procedures for handling dosim-
each particular application. Some of the principal ones are
eter materials during preparation must be developed to ensure
discussed individually as follows:
6.1.1 Self-Shielding of Neutrons—The neutron self-
4 shielding phenomenon oc
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