ASTM E844-09(2014)e1
(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)
SIGNIFICANCE AND USE
4.1 In neutron dosimetry, a fission or non-fission dosimeter, or combination of dosimeters, can be used for determining a fluence rate, fluence, or neutron spectrum in nuclear reactors. Each dosimeter is sensitive to a specific energy range, and, if desired, increased accuracy in a fluence-rate spectrum can be achieved by the use of several dosimeters each covering specific neutron energy ranges.
4.2 A wide variety of detector materials is used for various purposes. Many of these substances overlap in the energy of the neutrons which they will detect, but many different materials are used for a variety of reasons. These reasons include available analysis equipment, different cross sections for different fluence-rate levels and spectra, preferred chemical or physical properties, and, in the case of radiometric dosimeters, varying requirements for different half-life isotopes, possible interfering activities, and chemical separation requirements.
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 SI units are to be regarded as standard. Values 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.
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´1
Designation: E844 − 09 (Reapproved 2014)
Standard Guide for
Sensor Set Design and Irradiation for Reactor Surveillance,
E 706 (IIC)
This standard is issued under the fixed designation E844; the number immediately following the designation indicates the year of
original adoption or, in the case of revision, the year of last revision.Anumber in parentheses indicates the year of last reapproval.A
superscript epsilon (´) indicates an editorial change since the last revision or reapproval.
ε NOTE—Figures 1 and 2 were updated and editorial changes were made in September 2014.
1. Scope E1214Guide for Use of Melt Wire Temperature Monitors
for Reactor Vessel Surveillance, E 706 (IIIE)
1.1 This guide covers the selection, design, irradiation,
E2005Guide for Benchmark Testing of Reactor Dosimetry
post-irradiation handling, and quality control of neutron do-
in Standard and Reference Neutron Fields
simeters (sensors), thermal neutron shields, and capsules for
E2006GuideforBenchmarkTestingofLightWaterReactor
reactor surveillance neutron dosimetry.
Calculations
1.2 The values stated in SI units are to be regarded as
standard. Values in parentheses are for information only.
3. Terminology
1.3 This standard does not purport to address all of the
3.1 Definitions:
safety problems, if any, associated with its use. It is the
3.1.1 neutron dosimeter, sensor, monitor—a substance irra-
responsibility of the user of this standard to establish appro-
diated in a neutron environment for the determination of
priate safety and health practices and determine the applica-
neutron fluence rate, fluence, or spectrum, for example: radio-
bility of regulatory limitations prior to use.
metricmonitor(RM),solidstatetrackrecorder(SSTR),helium
accumulation fluence monitor (HAFM), damage monitor
2. Referenced Documents
(DM), temperature monitor (TM).
2.1 ASTM Standards:
3.1.2 thermal neutron shield—a substance (that is,
E170Terminology Relating to Radiation Measurements and
cadmium, boron, gadolinium) that filters or absorbs thermal
Dosimetry
neutrons.
E261Practice for Determining Neutron Fluence, Fluence
Rate, and Spectra by Radioactivation Techniques
3.2 Fordefinitionsorothertermsusedinthisguide,referto
E854Test Method for Application and Analysis of Solid
Terminology E170.
State Track Recorder (SSTR) Monitors for Reactor
Surveillance, E706(IIIB)
4. Significance and Use
E910Test Method for Application and Analysis of Helium
4.1 In neutron dosimetry, a fission or non-fission dosimeter,
Accumulation Fluence Monitors for Reactor Vessel
or combination of dosimeters, can be used for determining a
Surveillance, E706 (IIIC)
fluence rate, fluence, or neutron spectrum in nuclear reactors.
E1005Test Method for Application and Analysis of Radio-
Each dosimeter is sensitive to a specific energy range, and, if
metric Monitors for Reactor Vessel Surveillance, E 706
desired, increased accuracy in a fluence-rate spectrum can be
(IIIA)
achieved by the use of several dosimeters each covering
E1018Guide for Application of ASTM Evaluated Cross
specific neutron energy ranges.
Section Data File, Matrix E706 (IIB)
4.2 Awide variety of detector materials is used for various
purposes. Many of these substances overlap in the energy of
This guide is under the jurisdiction of ASTM Committee E10 on Nuclear
the neutrons which they will detect, but many different
Technology and Applicationsand is the direct responsibility of Subcommittee
materials are used for a variety of reasons. These reasons
E10.05 on Nuclear Radiation Metrology.
include available analysis equipment, different cross sections
CurrenteditionapprovedJune1,2014.PublishedJuly2014.Originallyapproved
fordifferentfluence-ratelevelsandspectra,preferredchemical
in 1981. Last previous edition approved in 2009 as E844–09. DOI: 10.1520/
E0844-09R14E01.
or physical properties, and, in the case of radiometric
For referenced ASTM standards, visit the ASTM website, www.astm.org, or
dosimeters, varying requirements for different half-life
contact ASTM Customer Service at service@astm.org. For Annual Book of ASTM
isotopes, possible interfering activities, and chemical separa-
Standards volume information, refer to the standard’s Document Summary page on
the ASTM website. tion requirements.
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States
´1
E844 − 09 (2014)
5. Selection of Neutron Dosimeters and Thermal Neutron tracks or helium) requires no time-dependent corrections and
Shields are therefore particularly valuable for long-term irradiations.
5.1.6 Fission detectors shall be chosen that have accurately
5.1 Neutron Dosimeters:
known fission yields. Refer to Method E1005.
5.1.1 The choice of dosimeter material depends largely on
5.1.7 In thermal reactors the correction for neutron self
the dosimetry technique employed, for example, radiometric
shielding can be appreciable for dosimeters that have highly
monitors, helium accumulation monitors, track recorders, and
absorbing resonances (see 6.1.1).
damagemonitors.Atthepresenttime,thereisawidevarietyof
5.1.8 Dosimeters that produce activation or fission products
detectormaterialsusedtoperformneutrondosimetrymeasure-
(that are utilized for reaction rate determinations) with half-
ments.Thesearegenerallyintheformoffoils,wires,powders,
lives that are short compared to the irradiation duration should
and salts. The use of alloys is valuable for certain applications
not be used. Generally, radionuclides with half-lives less than
such as (1) dilution of high cross-section elements, (2) prepa-
three times the irradiation duration should be avoided unless
rationofelementsthatarenotreadilyavailableasfoilsorwires
there is little or no change in neutron spectral shape or fluence
inthepurestate,and(3)preparationtopermitanalysisofmore
rate with time.
than one dosimeter material.
5.1.9 Tables 1-3 present various dosimeter elements. Listed
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
giesoftheprincipalresonancesarelistedinorderofincreasing
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 belowwhich5%oftheactivityisproducedandE above
for time differences between the end of the irradiation and the 05 95
which 5% of the activity is produced) for the U neutron
subsequentcounting.RefertoMethodE1005fornucleardecay
thermal fission spectrum are included.
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 0.51 to
decades for radiometric and SSTR dosimeters, 8 decades for
1.27 mm (0.020 to 0.050 in.) thick. However, because elemen-
HAFMs). Since fluence-rate levels at dosimeter locations can
tal cadmium (m.p. = 320°C) will melt if placed within the
range over 2 or 3 decades in a given experiment and over 10
vessel of an operating water reactor, effective thermal neutron
decades between low power and high power experiments, the
filters must be chosen that will withstand high temperatures of
proper sizing of dosimeter materials is essential to assure
light-water reactors. High-temperature filters include cadmium
accurate and economical analysis.
oxide (or other cadmium compounds or mixtures), boron
5.1.4 The estimate of radiometric dosimeter activity levels
(enrichedinthe Bisotope),andgadolinium.Thethicknessof
atthetimeofcountingincludeadjustmentsforthedecayofthe
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 Inreactors,feasibledosimeterstodatewhoseresponse
such calculations is (in the absence of fluence-rate perturba-
range to neutron energies of 1 to 3 MeV includes the fission
tions) as follows:
238 237 232
monitors U, Np, and Th. These particular dosimeters
2λt 2λt
1 2
A 5 N σ¯φα~1 2e !~e ! (1)
o
where:
TABLE 1 Dosimeter Elements—Thermal Neutron Region
A = expected disintegration rate (dps) for the product
Element of
Nuclear Reaction Available Forms
nuclide at the time of counting,
Interest
N = number of target element atoms, 10 7
o B B(n,α) Li B, B C, B-Al, B-Nb
59 60
φ = estimated fluence-rate density level,
Co Co(n,γ) Co Co, Co-Al, Co-Zr
63 64
Cu Cu(n,γ) Cu Cu, Cu-Al, Cu(NO )
σ¯ = spectral averaged cross section, 3 2
197 198
Au Au(n,γ) Au Au, Au-Al
α = product of the nuclide fraction and (if applicable)
115 116m
In In(n,γ) In In, In-Al
58 59
of the fission yield,
Fe Fe(n,γ) Fe Fe
-λt
1 54 55
Fe Fe(n,γ) Fe Fe
1−e = buildup of the nuclide during the irradiation
6 3
Li Li(n,α) H LiF, Li-Al
period, t ,
55 56
Mn Mn(n,γ) Mn alloys
-λt
e = decay after irradiation to the time of counting, t , 58 59 56
Ni Ni(n,γ) Ni(n,α) Fe Ni
and
Pu Pu(n,f)FP PuO , alloys
45 46
Sc Sc(n,γ) Sc Sc, Sc O
λ = decay constant for the product nuclide. 2 3
109 110m
Ag Ag(n,γ) Ag Ag, Ag-Al, AgNO
23 24
Na Na(n,γ) Na NaCl, NaF, NaI
5.1.5 ForSSTRsandHAFMs,thesametypeofinformation
181 182
Ta Ta(n,γ) Ta Ta, Ta O
2 5
as for radiometric monitors (that is, total number of reactions)
U (enriched) U(n,f)FP U, U-Al, UO ,U O , alloys
2 3 8
is provided.The difference being that the end products (fission
´1
E844 − 09 (2014)
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,α) H Li LiF, Li-Al
A 10 7
B(n,α) Li B B, B C, B-Al, B-Nb
A 58 59 56
Ni(n,γ) Ni(n,α) Fe Ni Ni
115 116m
1.457 In(n,γ) In In In, In-Al
181 182
4.28 Ta(n,γ) Ta Ta Ta, Ta O
2 5
197 198
4.906 Au(n,γ) Au Au Au, Au-Al
109 110m
5.19 Ag(n,γ) Ag Ag Ag, Ag-Al, AgNO
232 233
21.806 Th(n,γ) 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,γ) Co Co Co, Co-Al, Co-Zr
58 59
1038 Fe(n,γ) Fe Fe Fe
55 56
337.3 Mn(n,γ) Mn Mn alloys
63 64
579 Cu(n,γ) Cu Cu Cu, Cu-Al, Cu(NO )
3 2
0.2956243 Pu(n,f)FP Pu PuO , alloys
23 24
2810 Na(n,γ) Na Na NaCl, NaF, NaI
45 46
3295 Sc(n,γ) Sc Sc Sc, Sc O
2 3
54 55
7788 Fe(n,γ) 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
Low Median High Uncertainty
Reactions Interest Forms
A,C
E E E (%)
05 50 95
Np(n,f)FP Np 0.684 1.96 5.61 9.34 Np O , alloys
2 3
103 103m
Rh(n,n') Rh Rh 0.731 2.25 5.73 3.10 Rh
93 93m
Nb(n,n') Nb Nb 0.951 2.57 5.79 3.01 Nb, Nb O
2 5
115 115m
In(n,n') In In 1.12 2.55 5.86 2.16 In, In-Al
14 11
N(n,α) B N 1.75 3.39 5.86 — TiN, ZrN, NbN
U(n,f)FP U (depleted) 1.44 2.61 6.69 0.319 U, U-Al, UO ,U O , alloys
3 3 8
Th(n,f)FP Th 1.45 2.79 7.21 5.11 Th, ThO
9 6
Be(n,α) Li Be 1.59 2.83 5.26 — Be
47 47
Ti(n,p) Sc Ti 1.70 3.63 7.67 3.77 Ti
58 58
Ni(n,p) Co Ni 1.98 3.94 7.51 2.44 Ni, Ni-Al
54 54
Fe(n,p) Mn Fe 2.27 4.09 7.54 2.12 Fe
32 32
S(n,p) P S 2.28 3.94 7.33 3.63 CaSO ,Li SO
4 2 4
32 29
S(n,α) Si S 1.65 3.12 6.06 — Cu S, PbS
58 55
Ni(n,α) Fe Ni 2.74 5.16 8.72 — Ni, Ni-Al
46 46
Ti(n,p) Sc Ti 3.70 5.72 9.43 2.48 Ti
56 56 D
Fe(n,p) Mn Fe 5.45 7.27 11.3 2.26 Fe
56 53
Fe(n,α) Cr Fe 5.19 7.53 11.3 — Fe
63 60 E
Cu(n,α) Co Cu 4.53 6.99 11.0 2.36 Cu, Cu-Al
27 24
Al(n,α) Na Al 6.45 8.40 11.9 1.19 Al, Al O
2 3
48 48
Ti(n,p) Sc Ti 5.92 8.06 12.3 2.56 Ti
47 44
Ti(n,α) Ca Ti 2.80 5.10 9.12 — Ti
60 60 E
Ni(n,p) Co Ni 4.72 6.82 10.8 10.3 Ni, Ni-Al
55 54 F
Mn(n,2n) Mn Mn 11.0 12.6 15.8 13.54 alloys
A 235
Energy response range was derived using the ENDF/B-VI U fission spectrum, Ref (1), MT = 9228, MF = 5, MT = 18. The cross section and associated covariance
sources are identified in Guide E1018 and in Refs (2,3).
B
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 .
50 95 05
C
Uncertainty metric only reflects that component due to the knowledge of the cross section and is reported at the 1σ level.
D
Low manganese content necessary.
E
Low cobalt content necessary.
F
Low iron content necessary.
12 −2 −1 58
must be shielded from thermal neutrons to reduce fission 3×10 n·cm ·s to prevent significant loss of Co and
235 238 58m 3
product production from trace quantities of U, Pu, Co by thermal neutron burnout (4).
and Pu and to suppress buildup of interfering fissionable
238 238 237 6. Design of Neutron Dosimeters, Thermal Neutron
nuclides,forexample, Npand Puinthe Npdosimeter,
239 238 233 232
Shields, and Capsules
Pu in the U dosimeter, and Uinthe Th dosimeter.
Thermal neutron shields are also necessary for epithermal 6.1 Neutron Dosimeters—Procedures for handling dosim-
−7
spectrum measurements in the 5×10 to 0.3-MeV energy eter materials during preparation must be developed to ensure
range. Also, nickel dosimeters used for the fast activation
58 58
reaction Ni(n,p) Comustbeshieldedfromthermalneutrons
Theboldfacenumberinparenthesesreferstothelistofreferencesattheendof
in nuclear environments having thermal fluence rates above the guide.
´1
E844 − 09 (2014)
personnel safety and accurate nuclear environment character- diluted prior to analysis. The lower limit on dosimeter size
ization. During dosimeter fabrication, care must be taken in would be governed by a size that could be readily handled and
ordertoachievedesiredneutronfluence-rateresults,especially
would not be easily lost or overlooked.
in the case of thermal and resonance-region dosimeters. A
6.1.5 Temperature—In high-power reactor irradiations, do-
number of factors must be considered in the design of a
simetersmustbeconstructedtowithstandtheadverseenviron-
dosimetry set
...
This document is not an ASTM standard and is intended only to provide the user of an ASTM standard an indication of what changes have been made to the previous version. Because
it may not be technically possible to adequately depict all changes accurately, ASTM recommends that users consult prior editions as appropriate. In all cases only the current version
of the standard as published by ASTM is to be considered the official document.
´1
Designation: E844 − 09 E844 − 09 (Reapproved 2014)
Standard Guide for
Sensor Set Design and Irradiation for Reactor Surveillance,
E 706 (IIC)
This standard is issued under the fixed designation E844; 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 (´) indicates an editorial change since the last revision or reapproval.
ε NOTE—Figures 1 and 2 were updated and editorial changes were made in September 2014.
1. 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 SI units are to be regarded as standard. Values 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.
2. Referenced Documents
2.1 ASTM Standards:
E170 Terminology Relating to Radiation Measurements and Dosimetry
E261 Practice for Determining Neutron Fluence, Fluence Rate, and Spectra by Radioactivation Techniques
E854 Test Method for Application and Analysis of Solid State Track Recorder (SSTR) Monitors for Reactor Surveillance,
E706(IIIB)
E910 Test Method for Application and Analysis of Helium Accumulation Fluence Monitors for Reactor Vessel Surveillance,
E706 (IIIC)
E1005 Test Method for Application and Analysis of Radiometric Monitors for Reactor Vessel Surveillance, E 706 (IIIA)
E1018 Guide for Application of ASTM Evaluated Cross Section Data File, Matrix E706 (IIB)
E1214 Guide for Use of Melt Wire Temperature Monitors for Reactor Vessel Surveillance, E 706 (IIIE)
E2005 Guide for Benchmark Testing of Reactor Dosimetry in Standard and Reference Neutron Fields
E2006 Guide for Benchmark Testing of Light Water Reactor Calculations
3. Terminology
3.1 Definitions:
3.1.1 neutron dosimeter, sensor, monitor—a substance irradiated in a neutron environment for the determination of neutron
fluence rate, fluence, or spectrum, for example: radiometric monitor (RM), solid state track recorder (SSTR), helium accumulation
fluence monitor (HAFM), damage monitor (DM), temperature monitor (TM).
3.1.2 thermal neutron shield—a substance (that is, cadmium, boron, gadolinium) that filters or absorbs thermal neutrons.
3.2 For definitions or other terms used in this guide, refer to Terminology E170.
4. Significance and Use
4.1 In neutron dosimetry, a fission or non-fission dosimeter, or combination of dosimeters, can be used for determining a
fluence-rate, fluence rate, fluence, or neutron spectrum, or both, spectrum in nuclear reactors. Each dosimeter is sensitive to a
This guide is under the jurisdiction of ASTM Committee E10 on Nuclear Technology and Applicationsand is the direct responsibility of Subcommittee E10.05 on Nuclear
Radiation Metrology.
Current edition approved June 1, 2009June 1, 2014. Published June 2009July 2014. Originally approved in 1981. Last previous edition approved in 20032009 as
E844 – 03.E844 – 09. DOI: 10.1520/E0844-09.10.1520/E0844-09R14E01.
For referenced ASTM standards, visit the ASTM website, www.astm.org, or contact ASTM Customer Service at service@astm.org. For Annual Book of ASTM Standards
volume information, refer to the standard’s Document Summary page on the ASTM website.
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States
´1
E844 − 09 (2014)
specific energy range, and, if desired, increased accuracy in a fluence-rate spectrum can be achieved by the use of several
dosimeters each covering specific neutron energy ranges.
4.2 A wide variety of detector materials is used for various purposes. Many of these substances overlap in the energy of the
neutrons which they will detect, but many different materials are used for a variety of reasons. These reasons include available
analysis equipment, different cross sections for different fluence-rate levels and spectra, preferred chemical or physical properties,
and, in the case of radiometric dosimeters, varying requirements for different half-life isotopes, possible interfering activities, and
chemical separation requirements.
5. Selection of Neutron Dosimeters and Thermal Neutron Shields
5.1 Neutron Dosimeters:
5.1.1 The choice of dosimeter material depends largely on the dosimetry technique employed, for example, radiometric
monitors, helium accumulation monitors, track recorders, and damage monitors. At the present time, there is a wide variety of
detector materials used to perform neutron dosimetry measurements. These are generally in the form of foils, wires, powders, and
salts. The use of alloys is valuable for certain applications such as (1) dilution of high cross-section elements, (2) preparation of
elements that are not readily available as foils or wires in the pure state, and (3) preparation to permit analysis of more than one
dosimeter material.
5.1.2 For neutron dosimeters, the reaction rates are usually deduced from the absolute gamma-ray radioanalysis (there exist
exceptions, such as SSTRs, HAFMs, damage monitors). Therefore, the radiometric dosimeters selected must have gamma-ray
yields known with good accuracy (>98 %). The half-life of the product nuclide must be long enough to allow for time differences
between the end of the irradiation and the subsequent counting. Refer to Method E1005 for nuclear decay and half-life parameters.
5.1.3 The neutron dosimeters should be sized to permit accurate analysis. The range of high efficiency counting equipment over
which accurate measurements can be performed is restricted to several decades of activity levels (5 to 7 decades for radiometric
and SSTR dosimeters, 8 decades for HAFMs). Since fluxfluence-rate levels at dosimeter locations can range over 2 or 3 decades
in a given experiment and over 10 decades between low power and high power experiments, the proper sizing of dosimeter
materials is essential to assure accurate and economical analysis.
5.1.4 The estimate of radiometric dosimeter activity levels at the time of counting include adjustments for the decay of the
product nuclide after irradiation as well as the rate of product nuclide buildup during irradiation. The applicable equation for such
calculations is (in the absence of fluence-rate perturbations) as follows:
2λt 2λt
1 2
A 5 N σ¯ φα 12e e (1)
~ !~ !
o
2λt 2λt
1 2
A 5 N σ¯ φα~12e !~e ! (1)
o
where:
A = expected disintegration rate (dps) for the product nuclide at the time of counting,
N = number of target element atoms,
o
φ = estimated flux density level,
φ = estimated fluence-rate density level,
σ¯ = spectral averaged cross section,
α = product of the nuclide fraction and (if applicable) of the fission yield,
-λt
1 − e = buildup of the nuclide during the irradiation period, t ,
-λt
e = decay after irradiation to the time of counting, t , and
λ = decay constant for the product nuclide.
5.1.5 For SSTRs and HAFMs, the same type of information as for radiometric monitors (that is, total number of reactions) is
provided. The difference being that the end products (fission tracks or helium) requires no time-dependent corrections and are
therefore particularly valuable for long-term irradiations.
5.1.6 Fission detectors shall be chosen that have accurately known fission yields. Refer to Method E1005.
5.1.7 In thermal reactors the correction for neutron self shielding can be appreciable for dosimeters that have highly absorbing
resonances (see 6.1.1).
5.1.8 Dosimeters that produce activation or fission products (that are utilized for reaction rate determinations) with half-lives
that are short compared to the irradiation duration should not be used. Generally, radionuclides with half-lives less than three times
the irradiation duration should be avoided unless there is little or no change in neutron spectral shape or fluence rate with time.
5.1.9 Tables 1-3 present various dosimeter elements. Listed are the element of interest, the nuclear reaction, and the available
forms. For the intermediate energy region, the energies of the principal resonances are listed in order of increasing energy. In the
case of the fast neutron energy region, the 95 % response ranges (an energy range that includes most of the response for each
dosimeter is specified by giving the energies E below which 5 % of the activity is produced and E above which 5 % of the
05 95
activity is produced) for the U neutron thermal fission spectrum are included.
5.2 Thermal Neutron Shields:
5.2.1 Shield materials are frequently used to eliminate interference from thermal neutron reactions when resonance and fast
neutron reactions are being studied. Cadmium is commonly used as a thermal neutron shield, generally 0.51 to 1.27 mm (0.020
´1
E844 − 09 (2014)
TABLE 1 Dosimeter Elements—Thermal Neutron Region
Element of
Nuclear Reaction Available Forms
Interest
10 7
B B(n,α) Li B, B C, B-Al, B-Nb
59 60
Co Co(n,γ) Co Co, Co-Al, Co-Zr
63 64
Cu Cu(n,γ) Cu Cu, Cu-Al, Cu(NO )
3 2
197 198
Au Au(n,γ) Au Au, Au-Al
115 116m
In In(n,γ) In In, In-Al
58 59
Fe Fe(n,γ) Fe Fe
54 55
Fe Fe(n,γ) Fe Fe
6 3
Li Li(n,α) H LiF, Li-Al
55 56
Mn Mn(n,γ) Mn alloys
58 59 56
Ni Ni(n,γ) Ni(n,α) Fe Ni
Pu Pu(n,f)FP PuO , alloys
45 46
Sc Sc(n,γ) Sc Sc, Sc O
2 3
109 110m
Ag Ag(n,γ) Ag Ag, Ag-Al, AgNO
23 24
Na Na(n,γ) Na NaCl, NaF, NaI
181 182
Ta Ta(n,γ) Ta Ta, Ta O
2 5
U (enriched) U(n,f)FP U, U-Al, UO , U O , alloys
2 3 8
to 0.050 in.) thick. However, because elemental cadmium (m.p. = 320°C) will melt if placed within the vessel of an operating water
reactor, effective thermal neutron filters must be chosen that will withstand high temperatures of light-water reactors.
High-temperature filters include cadmium oxide (or other cadmium compounds or mixtures), boron (enriched in the B isotope),
and gadolinium. The thickness of the shield material must be selected to account for burnout from high fluences.
5.2.2 In reactors, feasible dosimeters to date whose response range to neutron energies of 1 to 3 MeV includes the fission
238 237 232
monitors U, Np, and Th. These particular dosimeters must be shielded from thermal neutrons to reduce fission product
235 238 239
production from trace quantities of U, Pu, and Pu and to suppress buildup of interfering fissionable nuclides, for example,
238 238 237 239 238 233 232
Np and Pu in the Np dosimeter, Pu in the U dosimeter, and U in the Th dosimeter. Thermal neutron shields are
−7
also necessary for epithermal spectrum measurements in the 5 × 10 to 0.3-MeV energy range. Also, nickel dosimeters used for
58 58
the fast activation reaction Ni(n,p) Co must be shielded from thermal neutrons in nuclear environments having thermal fluence
12 −2 −1 58 58m 3
rates above 3 × 103 × 10 n·cm n·cm ·s to prevent significant loss of Co and Co by thermal neutron burnout (4).
6. Design of Neutron Dosimeters, Thermal Neutron Shields, and Capsules
6.1 Neutron Dosimeters—Procedures for handling dosimeter materials during preparation must be developed to ensure
personnel safety and accurate nuclear environment characterization. During dosimeter fabrication, care must be taken in order to
achieve desired neutron fluxfluence-rate results, especially in the case of thermal and resonance-region dosimeters. A number of
factors must be considered in the design of a dosimetry set for each particular application. Some of the principal ones are discussed
individually as follows:
6.1.1 Self-Shielding of Neutrons—The neutron self-shielding phenomenon occurs when high cross-section atoms in the outer
layers of a dosimeter reduce the neutron flux fluence rate to the point where it significantly affects the activation of the inner atoms
of the material. This is especially true of materials with high thermal cross sections and essentially all resonance detectors. This
can be minimized by using low weight percentage alloys of high-cross-section material, for example, Co-Al, Ag-Al, B-Al, Li-Al.
It is not as significant for the fast region where the cross sections are relatively low; therefore, thermal and resonance detectors
shall be as thin as possible. Mathematical corrections can also be made to bring the material to “zero thickness” but, in general,
the smaller the correction, the more accurate will be the results. Both theoretical treatments of the complex corrections and
experimental determinations are published (5-1617,17).
6.1.2 Self-Absorption of Emitted Radiation—This effect may be observed during counting of the radiometric dosimeter. If the
radiation of interest is a low-energy gamma ray, an X ray, or a beta particle, the thickness of the dosimeter may be of appreciable
significance as a radiation absorber (especially for higher atomic number materials). This will lower the counting rate, which would
then have to be adjusted in a manner similar to that for the “zero thickness” correction in the case of self-shielding. Therefore, it
would again be desirable to use thin dosimeters in cases where the count rate is affected by dosimeter thickness. In the case of thick
pellets, it is usually possible to perform chemical separation of the radionuclide.
6.1.3 Fission Fragment Loss—It has been observed that fission foils of 0.0254-mm (0.001-in.) thickness lose a significant
fraction (approximately 7 %) of the fission fragments. Increasing the thickness to 0.127 mm (0.005 in.) will reduce this loss to
about 1 %.
6.1.4 Dosimeter Size:
6.1.4.1 The size of dosimeters and dosimetry sets is often limited by space available, especially in reactor applications where
volume in high fluxfluence-rate regions is very limited and in great demand for experimental samples. This fact, coupled with the
The boldface number in parentheses refers to the list of references at the end of the guide.
´1
E844 − 09 (2014)
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,α) H Li LiF, Li-Al
A 10 7
B(n,α) Li B B, B C, B-Al, B-Nb
A 58 59 56
Ni(n,γ) Ni(n,α) Fe Ni Ni
...
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