ASTM E844-09(2014)e2

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Designation: E844 − 09 (Reapproved 2014) E844 − 09 (Reapproved 2014)
Standard Guide for
Sensor Set Design and Irradiation for Reactor Surveillance,
E 706 (IIC)Surveillance
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.
ε NOTE—The title and Referenced Documents were udpated in May 2017.
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.
1.4 This international standard was developed in accordance with internationally recognized principles on standardization
established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued
by the World Trade Organization Technical Barriers to Trade (TBT) Committee.
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
E910 Test Method for Application and Analysis of Helium Accumulation Fluence Monitors for Reactor Vessel Surveillance,
E706 (IIIC)Surveillance
E1005 Test Method for Application and Analysis of Radiometric Monitors for Reactor Vessel Surveillance
E1018 Guide for Application of ASTM Evaluated Cross Section Data File, Matrix E706 (IIB)File
E1214 Guide for Use of Melt Wire Temperature Monitors for Reactor Vessel Surveillance, E 706 (IIIE)Surveillance
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.
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, 2014. Published July 2014. Originally approved in 1981. Last previous edition approved in 2009 as E844 – 09. DOI:
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
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E844 − 09 (2014)
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, 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.
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 fluence-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
where:
A = expected disintegration rate (dps) for the product nuclide at the time of counting,
N = number of target element atoms,
o
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:
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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
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
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 × 10 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 fluence-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 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 (5-17).
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:
The boldface number in parentheses refers to the list of references at the end of the guide.
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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
1.4
...