Assessment of Gamma Dose Rate for Hypothetical Radioactive Waste Container

Metallic solid radioactive waste class low-intermediate short lived level waste (LILSL RW) is the main type of radioactive waste generated from decommissioning operations. Transport, storage and disposal regulations require for gamma emitting radioactive waste (mainly by 137Cs isotope), that the dose rate in the proximity of the container should stand below a certain threshold. Also, the conditioning technique (using cementation technique) based on certain matrix with specific ratios should be able to attenuate the gamma radiation activity to the minimum level or to acceptable dosage rate at distance of 1 m from the container. In this paper, in absence of suitable labs for waste package assessment, hypothetical method present to assess dose rate in safe way, assumption based on metallic waste pieces contaminated with (137Cs) were conditioned with cement matrix and contained in carbon steel drum volume 220 L, 60 cm diameter then dose rate measurement applied in vicinity of the container. Instead of real contaminated metallic waste (137Cs, D∘=20 mR/h), gamma radioactive point source was positioned in different places in front of cross section of the cemented free metallic waste and gamma dose rates were measured on the outer side of the drum sample using NaI detector dose meter device. Readings showed good attenuation of gamma radiation activity (low dose rates), efficiency of the cement matrix to decrease the dose rate of (137Cs, 0.662 Mev) gamma radiation lower to acceptable values and with waste acceptance criteria and regulation.


Introduction
Radioactive waste handling activities are hazardous as concern for both contamination and external exposure; therefore, strict regulations are applied for radiation protection in this field [1]. Dose rate in the vicinity of waste drum has to be kept below certain constraints throughout entire treatment and conditioning process and for the final storage; the dose rate should meet Waste Acceptance Criteria (WAC). Many experiments and tests applied to conditioned radioactive waste drum to check their purposes such as quality of encapsulation process or evaluation the attenuation of gamma activity to minimum level, also to determine and assess the dose rate for conditioned RW drums to storage or disposal [2]. Here one of the test that should be taken in RW management include cut open the package RW (cemented drum) in horizontal direction in safe-secured conditions to examine the enteric of the cemented RW drum and for gamma attenuation determination (shielding function), the

Abstract
Metallic solid radioactive waste class low-intermediate short lived level waste (LIL-SL RW) is the main type of radioactive waste generated from decommissioning operations. Transport, storage and disposal regulations require for gamma emitting radioactive waste (mainly by 137 Cs isotope), that the dose rate in the proximity of the container should stand below a certain threshold. Also, the conditioning technique (using cementation technique) based on certain matrix with specific ratios should be able to attenuate the gamma radiation activity to the minimum level or to acceptable dosage rate at distance of 1 m from the container. In this paper, in absence of suitable labs for waste package assessment, hypothetical method present to assess dose rate in safe way, assumption based on metallic waste pieces contaminated with ( 137 Cs) were conditioned with cement matrix and contained in carbon steel drum volume 220 L, 60 cm diameter then dose rate measurement applied in vicinity of the container. Instead of real contaminated metallic waste ( 137 Cs, D ∘ =20 mR/h), gamma radioactive point source was positioned in different places in front of cross section of the cemented free metallic waste and gamma dose rates were measured on the outer side of the drum sample using NaI detector dose meter device. Readings showed good attenuation of gamma radiation activity (low dose rates), efficiency of the cement matrix to decrease the dose rate of ( 137 Cs, 0.662 Mev) gamma radiation lower to acceptable values and with waste acceptance criteria and regulation.

International Journal of Applied Science -Research and Review
ISSN 2394-9988 incorporate as much activity as possible to optimize these cots. In addition, many studies could be followed in this field for labs poor in Rad waste drums characterization. The estimation of gamma dose for homogenous waste containers are widely field because of variety of radioactive waste materials for each country and with variety of their activities [3].

Methods
To achieve the objectives of the present study, Iraq Portland cement and additives are prepared and mixed in certain ratios as Tables 1 and 2 show the specification of cementation matrix ratios while Table 3 represent the chemical analysis of container alloy components that used in the waste management facility.
The cement matrix and additives was mixed for 30 min, which were considering enough to achieve good homogeneity. The mixture density (ρ) about 1.675 g/cm 3 dropped into the drum where the clean solid waste was collected in iron basket and centered in the container and left for few moment on the vibrate stage to let gas bubbles escaped from the surface. The sample left for 28 days where the solidification of the cement-waste mixture is completed. Figure 1 show conditioned free waste package after cutting process using electrical machine to get half shape pieces. For gamma dose rate validation, a sample of half-half package waste form was prepared and head to 137 Cs gamma source to assumed the waste has been exposed to such gamma radiation rays from many distance (R) along the bulk diameter and that head to different thickness (x) or depth has gamma radiation penetrate through cement matrix shield and reach the detector window. The detection system was dose meter of (NaI) crystal type Ludlum held device with 40% efficiency of gamma-ray 0.661 Mev, background with 9 µR/h and used at contact with the outer side of waste package. Dose rate measurement and apparatus design in Figure 2, which represents schematic representation of method design.

Results
The well-known photon linear attenuation coefficient (µ) or shielding equation may be calculated using equation below: Where A ͦ the incident gamma-ray activity and D ∘ dose rate, which obtained without inserting any sample between the detector and the source and A, D, when the incident photons obtained for the cement waste bulk of thickness (x). The attenuation coefficient µ could be calculated in terms of dose rate and activity terms using narrow collimated mono-energetic beam of 137 Cs gamma source. The linear attenuation coefficient µ assumed to be obtained by the following formula: Low magnification optical graphs of cut open conditioned solid waste container.    Table 3 The chemical analysis of container waste alloy using AAS technique. (2) The dose rates and attenuation coefficient µ were performed for cemented waste cross section along diameter (R) in relative to their thickness (x). While µ depend on density of absorber material and cross sections of gamma-ray reactions with absorber material readings show ratios of errors as was expected. By assume equation (2)  . Figure 3 show discrepancies in the results, which are mainly due to internal in homogeneity of the waste form, which can alter the dose rate measurement results, it is to be noted that the main hypothesis of this estimation is the homogeneity of the waste form content and the uniformity of the activity distribution. By comparing the dose Schematic representation of dose rate measurement to waste container.  Relation of ln (D ∘ /D) against cemented waste specimen depth (thickness).

Figure 3
Variation of final Dose rate D relate to 137 Cs gamma source location along the cement waste cross section diameter, R (0-52 cm).

Figure 4
Linear attenuation coefficient (µ) against shield thickness (x). Figure 5 International Journal of Applied Science -Research and Review ISSN 2394-9988 Relation between ln A ∘ /A and thickness shield(x).

Figure 6
Linear attenuation coefficient (µ) and thickness shield(x).   3. The minimum dose rate values was in contact of waste container bulk exposed to 137 Cs gamma source was (3.11 µSv/h) and the maximum dose values was (7.62 µSv/h) related to the thickness of the penetration depth of the gamma radiation 29.5 and 19.5 cm.
4. The results proved that the cementation matrix (ratios) and the iron basket with the carbon steel container metals have an impact on attenuation performance of gamma radiation beside the bulk density and porosity level of the solidified cement-waste mix.
5. Substantial improvement of about (28.86 and 20.73%) in attenuation performance in term of dose rate ratios at R (5 and 30) cm for initial dose rate D ∘ (26.4 and 15 µSv/h) in air of cement waste sample depth was attained for solid waste and using Iraqi Portland cement local product.
6. Solid waste conditioning and treatment utilized successfully due to the structure of open cut cross section of cement waste bulk.
2. The maximum linear attenuation coefficient in terms of dose rate or activity were attained for cement matrix ratios incorporate with iron basket for 15-20 % waste, 80-85% cement matrix for 137 Cs gamma source of initial dose 15 µSv/h was 0.053 1/cm.