Answer to Question #8308 Submitted to "Ask the Experts"

Category: Instrumentation and Measurements — Surveys and Measurements (SM)

The following question was answered by an expert in the appropriate field:

Q

I was wondering how you would determine dose or what type of dosimeter you could use in order to determine dose and/or dose rate in a reactor core. I know that you are dealing with high doses and many common dosimeters (e.g., TLD) would be saturated and activated. Also, is there any possible way to separate the dose contribution due to different fields (i.e., neutron and gamma irradiation). Can you use LiF or CR-39?

A

As you are aware, dosimeters for assessment of reactor in-core doses are challenged by a number of factors, including the physical environment, which may include water, likely at elevated temperature and possibly at elevated pressure, depending on the type of reactor of interest and whether you wish to make measurements while the reactor is operating or following shutdown. Additionally, the often high and complex radiation field puts severe demands on the dosimetry devices. High-intensity gamma radiation fields with potential dose rates exceeding 2.78 Gy s-1 make some common dosimetry devices unsuitable for any reasonable duration exposures. When the reactor is running, high fluence rates of thermal and fast neutrons often exceed 1014 cm-2 s-1 and represent water absorbed dose rates exceeding 110 Gy s-1 for thermal neutrons and possibly 1 to 2 orders of magnitude greater for fast neutrons. Naturally, when the reactor is shut down neutron dose rates will drop by several orders of magnitude but may still be appreciable, depending on the operating history of the reactor and the buildup of various transuranic species in the fuel matrix.

In general, it is easier to make measurements in the core of many research reactors than it is in large power reactors. Research reactor fuel is often much more easily accessible for manipulation, and the reactors are often designed so that various samples can be introduced into selected core locations. Most research reactors also operate at a small fraction of the power of commercial power reactors and operate at much lower temperatures and often at no added pressure beyond the head of water that may cover the core. If the reactor is shut down so that dosimeters can be reasonably quickly put into the desired core location(s) and irradiated for a relatively short time at reasonably low temperature, it is possible that some TLDs can be used for gamma measurements. For example, it has been shown that the TL material Al2O3(Mg:Y) exhibits a linear response up to about 1,000 Gy, which is likely sufficient for some post-shutdown in-core measurements. If gamma dose rates have been reduced appreciably through post shutdown decay, some other more common TLDs may be useful. For example, CaF2(Mn), also known as TLD-400, and Li2B4O7(Mn), also known as TLD-800, each provide an acceptable response up to about 100 Gy. The most common LiF (Mg:Ti), TLD-100, is limited because of its lack of linearity and rather restricted range even in the supralinear region, although the more recent LiF(Cu:Mg:P) elements are linear to about 18 Gy and can be used in a sublinear response region up to about 60 Gy or somewhat more. Global Dosimetry Systems has done developmental work on a high-dose TLD, using TLD-100 and TLD-700 (depleted in 6Li), that was intended to achieve doses of about 1,000 Gy; this has been discussed in a paper by Sander Perle in Radiation Protection Dosimetry. There has also been some use of pure LiF in an optical absorption mode, measuring optical density increases (between about 200 nm and 1,100 nm) associated with high doses.

A variety of other types of dosimetry devices have been used for high-dose measurements in various environments, especially in high-dose product processing. These include various glass-based dosimeters. Some glasses are used as radioluminescent devices, but higher doses can often be measured using optical absorption measurements in which the change in coloration of the glass is assessed by measuring light transmission through the glass; this technique can apply up to about 10 kGy with some commercial glasses. Some plastics, such as Perspex, an acrylic plastic, have been used for many years for high-dose measurements. Changes in coloration of the plastic following irradiation are measured, usually by measuring light transmission over restricted wavelength intervals. Doses to somewhat beyond 50 kGy can be measured by this technique. A variety of liquid chemical solution dosimeters have been used for high-dose measurements. One such is the ceric sulfate dosimeter in which the reduction of Ce-4 to Ce+3 ion, brought about through irradiation, is measured and correlated with gamma dose; this dosimeter is useful to roughly 60 kGy. A number of dye systems, both liquid and solid, that respond by changing color upon irradiation, have been used for gamma dose measurements. Chemical systems, and especially liquid systems (vs. solid), have the complication that modest changes in temperature can significantly alter the chemical transformation rate so that corrections may be required if the system is used at elevated temperatures. Another powerful dosimetric technique for high-dose measurements is electron spin resonance (ESR), in which magnetic resonance measurements are made to assess the population of species with unpaired electrons that are produced in selected materials following irradiation. The most popular material used has been alanine, which has been used for measuring gamma doses to about 100 kGy. ESR techniques require specialized and quite expensive equipment, and this has restricted its use for many facilities.

Many of these techniques, as well as others that we have not mentioned, have been discussed in the literature. If you are interested in more detail on these, you can search the Internet and refer to some existing references from IAEA; one is High Dose Dosimetry for Radiation Processing, Proceedings of a Symposium, the Second in its Field, Vienna, 5–9 November 1990 (STI/PUB/846), 1991, and a couple of other related documents are IAEA Tec-Doc 1156, Dosimetry for Radiation Processing and IAEA Tec-Doc 1070, Techniques for high dose dosimetry in industry, agriculture and medicine.

All of the techniques noted above for gamma dose measurement have some potential for also responding to neutrons, and this can complicate dose interpretation. Response to thermal neutrons can be minimized by using materials with small thermal neutron cross sections—e.g., if Li-based TLDs are employed, use dosimeters depleted in the 6Li isotope (also depleted in 10B if boron is present). Hydrogenous systems may produce some response to fast neutrons through recoil proton production, and this may require evaluation in order to judge its significance and make corrections if necessary.

Techniques for assessment of neutron doses using dosimeters are restricted at very high doses. Techniques using track etch methods, such as afforded by CR-39, and other techniques such as the very sensitive bubble dosimeters are generally too sensitive for use at the very high neutron dose rates that might prevail in an operating reactor core. They might have application in cases when the reactor has been shut down and neutron dose rates are low. A method that has been widely applied to neutron fluence rate determinations in operating reactors has been to use activation threshold detectors. By proper selection of a variety of different materials that undergo specific neutron interactions with neutrons above respective threshold energies to produce radioactive products, neutron fluence rates of various energy neutron groups may be determined. Respective dose rates may be determined through appropriate fluence-to-dose conversions. Some examples of neutron threshold reactions and their respective effective energy thresholds are 237Np(n,f)FP (0.75 MeV) (f refers to fission and FP to fission products); 238U(n,f)FP (1.45 MeV); 32S(n,p)32P (3 MeV); and 27Al(n,a)24Na (8.1 MeV). Thermal neutron fluence rates are generally determined through thermal neutron activation of selected materials—e.g., 64Cu produced by thermal neutron capture by 63Cu. In research reactors it is fairly common to map the thermal neutron fluence rates in the core by inserting long copper wires alongside the fuel elements, operating at low power, removing the wires, and measuring the 64Cu along the length of the wires to evaluate the longitudinal variation in thermal neutron fluence rate. All of the neutron measurement techniques have the advantages that they are not influenced by gamma radiation and are unaffected by temperature changes.

The particular dosimeters that are best suited for a given application will depend on the specific conditions under which the measurements must be made, and the process for obtaining good dose (rate) measurements can be challenging and  sometimes tedious. In addition to passive dosimetric devices, there are also various types of hard-wired instrumentation, not discussed here, that have been designed for in-core use and can provide important information useful in assessing neutron and gamma intensities in the operating reactor.

I hope this is helpful.

George Chabot, PhD, CHP

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