SAMPLING AND RADIOLOGICAL ANALYSIS OF COMPONENTS OF THE TRIGA REACTOR AT THE MEDICAL UNIVERSITY OF HANNOVER

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1 SAMPLING AND RADIOLOGICAL ANALYSIS OF COMPONENTS OF THE TRIGA REACTOR AT THE MEDICAL UNIVERSITY OF HANNOVER ABSTRACT Gabriele Hampel, Heinrich Harke, Medical University of Hanover, Germany Uwe Klaus, Wieland Kelm, Babcock-Noell Nuclear GmbH, Germany Gunter Lörcher, TESSAG-NIS Ingenieurgesellschaft, Germany In preparation for dismantling the facility a number of samples were taken from the various reactor components in The aim of the samples being taken was to establish the radiological condition of the facility in more detail, in particular the condition of the activated components in the reactor tank and the biological shield in the core area. This paper describes the taking of the samples, the results of the radiological and chemical analyses and the preliminary calculations of activations of reactor components. INTRODUCTION After shutting down the facility at the end of 1996 the spent TRIGA fuel elements from the research reactor at the Medical University of Hanover (MHH) were returned to the United States in the summer of 1999 and thus disposed of for the MHH. Consequently one of the main prerequisites for dismantling the TRIGA reactor as planned has been fulfilled. Up to now the calculated estimates for the reactor components had been based mainly on the details provided in the facility documentation when operation started at the beginning of the 1970s, showing that the evaluation of the activity and dose rates was too high. This was confirmed in 1998 in the course of measuring contamination and dose rates when samples were taken from some reactor components before the fuel elements were removed. For example, drill samples were taken from the bottom part of a graphite blind element and from the central radiation beam tube in the core area and then analyzed by the U.R.A. Laboratory of the University of Regensburg. TAKING THE SAMPLES In general The samples were taken with permission from the responsible atomic energy authority on the basis of specialized instructions including a plan for carrying out each step. Individual work plan were drawn up and approved for each of the measures. The specialized instructions include general information about the sampling plan, accountability and how to take the samples as well as about radiation protection and safety evaluation. The sampling plans describe each step of the procedure including the required tests and checks. The measures required for guaranteeing work safety, fire protection and radiation protection are set down in each work plan.

2 In order to obtain the sample material, three charges each of sufficient amounts were taken from the corresponding samples as follows: one sample for chemical and radiological evaluation by an external, certified analysis center one sample for revision by an expert from the atomic energy authority one sample in reserve for the MHH The sample material was placed in a suitable receptacles. Samples were taken from the following materials and components from areas of high neutron flux (see figure 1): graphite from a graphite blind element aluminum from the reactor tank aluminum from the top grid plate stainless steel from the screw in the top grid plate Baryt concrete and reinforcement irons from the biological shield Radiation protection measurements were made while each of the samples were being taken. A description of how the samples were taken from the above mentioned components and the results of the radiation protection measurements follows. Graphite blind element Sampling The graphite blind element with the highest dose rate at a distance of half a meter (20 µsv/h) was chosen for the graphite sample. The graphite blind element was pulled out of the reactor tank with the fuel element gripper, placed in a special shutter facility and then in a ventilated sample box where the samples were extracted. The spent air from the box was fed into the reactor hall spent air ventilation system through an active carbon filter. The graphite blind element was drilled a total of 15 times by using a hand drill with a spiral drill. In addition, pieces of graphite to be used for determining the H-3 content were chipped off between the drill holes by using a hammer and a chisel (see figure 2). Altogether approx. 47 g of graphite powder and 13 g of graphite chips were extracted for the sample. Radiation protection The maximum contact dose rates were: 280 µsv/h on the shutter container near the drill hole

3 50 µsv/h on the surface of the sample box 25 µsv/h on the surface of the sample receptacle with 19 g of graphite powder The maximum dose per person due to taking samples from the graphite blind element was 12 µsv. Wall of the reactor tank Sampling Samples were taken from the wall of the reactor tank above the water level to the left and right of the bridge by using a hand drill with a spiral drill. The wall of the tank was drilled a total of at least 15 times on each side to a depth of about 5 mm (wall thickness 7 mm). Altogether approx. 30 g of aluminum shavings were extracted for the sample. Radiation protection The contact dose rate at the wall of the reactor tank and on the surface of the containers with the samples corresponded to the values in the foundation of the reactor hall, 0.08 µsv/h. The maximum dose per person due to taking samples from the wall of the reactor tank was 1 µsv. Top grid plate Sampling First the top grid plate was cleaned with a wire brush at those places where the samples were to be taken, then the top grid plate (about 19 mm thick) was drilled a total of 13 times by using a hand drill with a drill shaft with a spiral drill without drilling completely through the plate. Due to the short length of the shavings it was necessary to drill more often than had been planned in order to obtain the required amount for the sample. Altogether approx. 21 g of aluminum shavings were extracted for the sample. Radiation protection A maximum contact dose rate of 17 µsv/h was measured on the surface of the sample receptacle. The maximum dose per person due to taking samples from the top grid plate was 1 µs v. Steel screw Sampling The top grid plate is fastened to the reflector in the reactor tank by 4 stainless steel screws. In order to take the sample one of these screws was removed with a special rod tool and placed in the shutter facility outside the reactor tank in the reactor room. Then some shavings were removed by using a hand drill with a spiral drill.

4 Approx. 10 g of steel shavings were obtained and placed in sample receptacles with a lead shield. The steel screw, which was removed from the grate was replaced by a new one. Radiation protection The maximum dose rates were: 360 µsv/h over the reactor tank at a distance of 50 cm from the unshielded steel screw 1.8 µsv/h at the drill opening of the shielded container with the screw inside approx. 1 µsv/h in the work area while drilling the screw 30 µsv/h on the surface of the lead shield of the sample receptacle. The maximum dose per person due to taking the sample from the steel screw was 7 µsv. Biological shield Sampling In order to obtain the samples from the concrete both axial and radial drillings were made in the biological shield by using a concrete drill (drill diameter 72 mm). A total of 10 radial drillings and one axial drilling were made. The radial drillings stopped just short of the outer edge of the wall of the reactor tank. The axial drilling was made at a distance of approx. 60 mm from the wall of the reactor tank (the edge of the drill hole) to a depth of approx. 6.5 m. The water required to rinse the drill head was channeled along the drill rod in a closed cycle consisting of a storage tank, circulation pump, spent water adapter on the drill and suction sleeve on the drill rod. The concrete sludge was collected in the storage tank (see figure 3). Samples from the drillings were obtained by using hammer and chisel, which were then packed in foil, marked and stored in U-profiles. The radial drillings in the biological shield were made near the horizontal radiation beam tube and to the left at a 90 angle. The depth of the drillings was approx. 1.7 m. The wall of the reactor tank was not damaged, just slightly scratched. The material extracted from each drilling was less than 1 m long. Reinforcement irons with a maximum diameter of 20 mm were found during the drillings. Separate shavings samples were taken from the reinforcement irons. Due to an unexpectedly high density of reinforcement irons it was only possible to carry out one axial drilling approx. 140 to the right of the horizontal radiation beam tube in the biological shield. A special probe was used to monitor the course of the drilling meter by meter in order to keep the safety distance to the reactor tank to a minimum. At the end of the drill hole the distance between the wall of the tank and the drilling was approx. 50 mm. The required sample material was obtained analogously to the radial drillings.

5 A total of approx. 5,500 g of Baryt concrete and approx. 1,000 g of reinforcement irons were obtained as samples. Radiation protection The highest doses during the radial drillings were measured near the radiation beam tube on a level with the bottom grid plate as follows: 180 µsv/h at the wall of the reactor tank 25 µsv/h approx. 25 cm in front of the wall of the reactor tank 0.15 µsv/h on the outside of the drill hole The following maximum dose rates were measured at the drilling positions at a 90 angle to the radiation beam tube: 3 µsv/h at the wall of the reactor tank 0.5 µsv/h approx. 25 cm in front of the wall of the reactor tank µsv/h on the outside of the drill hole The dose rate at the beginning of the drill hole during the axial drilling was consistent with the base value of 0.08 µsv/h. The maximum dose per person while samples were being extracted from the biological shield was 22 µsv during 26 working days, i.e. an average of 1 µsv/day. Results of radiological monitoring The samples were taken during a total of 32 working days. Of the 5 persons who did the work, the overseer was involved constantly and the other 4 persons from the reactor operation staff were involved intermittently. The personal dose for the overseer amounted to a total of 43 µsv, the dose for the reactor operation staff approx. 98 µsv. Thus the collective dose that resulted was approx. 141 µsv. The rinse water and the drill sludge were collected in barrels and samples were taken. A total of approx. 155 liters of rinse water (Ba-133 activity less than 5 Bq/l) and approx. 266 liters of drill sludge (maximum activity for Ba-133 approx. 2.5 kbq/l, for approx. 0.7 kbq/l and for approx. 0.2 kbq/l) resulted from the drillings. The radiological analysis of the filters in the ventilation control system of the reactor room, which were changed each working day, indicated only radionuclides of natural origin. No radioactivity was discharged into the spent air system.

6 EVALUATION OF THE SAMPLES In general The samples were taken from components of different materials characteristic for the reactor facility and from easily accessible areas in order to provide the best possible survey of the chemical composition and the radiological data. The current radiological condition at particular points is to be determined with these samples in order to make it possible to provide a current estimate of the existing inventory of activity in the reactor facility. At the same time the results of the analysis will serve as comparative values for calculating the activity specific to the components as planned. In particular samples were taken from the drilled concrete at different points, e.g. directly on the outer side of the wall of the reactor tank. Table I provides an overview of the chemical and radiological analyses carried out. Table I. Chemical and radiological analyses Material Radionuclides Stable elements Graphite C-14, H-3 Co, Fe Aluminum, Ez-152 Co, Zn Stainless steel Fe-55, Ni-63, Fe, Co, Ni, Mn Baryt concrete H-3, C-14, Ba-133,, Co, Fe, Ni, Eu, Ba, Ca Reinforcement irons in the Baryt concrete Fe-55, Ni-63, Fe, Co, Ni, Mn A total of 2 graphite samples, 3 aluminum samples, one stainless steel sample, 22 Baryt concrete samples and 2 reinforcement iron samples were sent to VKTA Rossendorf for external analysis. Radiological measurements for comparison purposes were carried out at the MHH for the graphite, aluminum and stainless steel. Analyses results The samples were prepared and analyzed at the VKTA Rossendorf Laboratory for Environmental and Radionuclide Analyses. The laboratory is accredited through the Deutsche Akkreditierungssystem Prüfwesen GmbH in accordance with regulation DIN EN The samples were handed over to the laboratory in the form of shavings (aluminum, stainless steel and reinforcements irons), as drill powder (graphite) and pieces (concrete). In order to determine the H-3 content in the graphite and concrete, samples were extracted without the introduction of heat into the sample material. The samples were prepared at the VKTA in accordance with the regulations pertaining to radiological and chemical analyses and were then analyzed by the following methods: high resolution gamma spectrometry with HP germanium detector (type n) comprehensive beta measurement (low level alpha beta counter)

7 liquid scintillation spectrometry after radio-chemical separation for determining H-3, C- 14, Fe-55 and Ni-63 mass spectrometry with inductively coupled plasma (ICP-MS) for determining traces of chemical elements The results are summarized in Tables II to V. The essential radionuclides and traces of chemical elements are shown in each table. The amounts necessary for ascertaining their presence were determined in accordance with regulation DIN

8 Table II. Results of the radiological analyses of the reactor components Nuclide Components Material Place where sample was taken Graphite blind element Graphite blind element (1998)** Graphite Al1100F Drill sample and pieces from the entire length of the element Drill sample from the bottom part Top grid plate AlMg3 F18 Shavings from the outer area Central radiation beam tube (1998)** Steel screw AlMg3 F18 Stainless steel Drill sample from the core area Fastening screw from the top grid plate H-3 C-14 Fe-55 Ni-63 Zn-65 Zn-65 Cs-134 Fe-55 Zn-65 Fe-55 Ni-63 Specific activity in Bq/g 29,000 ± 4,000 5,200 ± 700 2,500 ± 400 7,000 ± 1,200 1,200 ± ± 2 105,600 ± 8,500 7,480 ± 150 1,100 ± ± ± 15 8,100 ± 1,300 < 40 < ± ± ,000 ± 14,000 60,180 ± 1,180 3,480 ± ± 17 < 40,000 3,100,000 ± 400,000 11,000,000 ± 1,000,000 2,800,000 ± 300,000 Reinforcement irons Steel St cm from the outside wall of the tank on a level with the top grid plate Fe-55* Ni-63* *Analyses results from the MHH **Analyses results from the U.R.A at the University of Regensburg 2,800 ± ± 40 3 ± ± 0.26

9 Table III. Results of the radiological analyses of the Baryt concrete in the biological shield Horizontal position Place where sample was taken Nuclide Specific Activity in Bq/g Vicinity of the radiation beam tube Vicinity of the radiation beam tube Vicinity of the radiation beam tube At a 90 angle to the left of the radiation beam tube At a 90 angle to the left of the radiation beam tube At a 140 angle to the right of the radiation beam tube On the outside wall of the tank on a level with the bottom grid plate 5 cm in front of the outside wall of the tank on a level with the bottom grid plate 50 cm in front of the outside wall of the tank on a level with the bottom grid plate 5 cm in front of the outside wall of the tank on a level with the middle of the core 30 cm in front of the outside wall of the tank on a level with the middle of the core 6 cm in front of the outside wall of the tank on a level with the middle of the core H-3 Ba-133 Cs-134 H-3 Ba-133 Cs-134 H-3 Ba-133 Cs-134 H-3 Ba-133 Cs-134 H-3 Ba-133 Cs-134 H-3 Ba-133 Cs-134 Not sampled 21 ± ± ± ± ± ± ± ± ± ± ± 0.7 Not sampled < ± 0.2 < ± 0.05 < ± ± ± 0.5 < ± ± 0.02 Not sampled < ± 0.12 < ± 0.02 < ± 0.16 < ± 0.15 < ± 0.03 < 0.034

10 Table IV. Results of the chemical trace analyses of the reactor components Element Component Material Place where sample was taken Graphite blind element Graphite Drill sample and pieces from the entire length of the element Top grid plate AlMg3 F18 Shavings from the outer area Steel screw Stainless steel Fastening screw from the top grid plate Fe Co Eu Co Zn Sr Eu Mn Fe Co Ni Mass of element trace/ total mass mg / kg 1,290 ± ± ± ± ± ± 0.4 < ,000 ± 1, ,000 ± 80,000 2,700 ± ,000 ± 12,000 Table V. Results of the chemical trace analyses of the Baryt concrete and reinforcement irons Component Material Place where sample was taken Element Mass of element trace/ total mass mg / kg Reactor shield Baryt concrete On the outside wall of the tank on a level with the bottom grid plate, next to the radiation beam tube Reactor shield Baryt concrete At a 90 angle to the left of the radiation beam tube, 30 cm in front of the outside wall of the tank on a level with the middle of the core Reinforcement irons Steel St cm from the outside wall of the tank on a level with the top grid plate Ca Fe Co Ni Ba Eu Ca Fe Co Ni Ba Eu Mn Fe Co Ni 42,000 ± 4,000 4,000 ± ± ± ,000 ± 40, ± 40 42,060 ± 5,000 4,000 ± ± ± ,000 ± 50, ± 40 8,500 ± ,000 ± 100, ± ± 30

11 Results of the preliminary calculations With the help of the NIS company s AKAT program the specific activity for essential components of the reactor and the biological shield were estimated. The calculations were based on the following assumptions: The reactor was in operation from April 1973 to May 1996 at a maximum power level of 250 kw. The decay time was assumed to be 4 years. The thermal and fast neutron flux values quoted by the General Atomic company in 1964 were used. Details of the material used in the components were taken from the documentation if it existed and supplemented with the results of the chemical analysis (see Tables IV and V). If there were no details available the literature and experience from other reactor facilities were taken as a basis. The neutron cross-sections were compiled by the NIS company from various literature sources. In Table VI there are some examples of specific activity determined from the AKAT calculations. For the purpose of comparison the last column shows the relation between the calculated results and those of the sample analysis. Table VI. Results of the AKAT calculations Component Nuclide Specific Activity in Bq/g Graphite blind element H-3 C-14 Bottom part of the graphite blind element Aluminum from the top grid plate Central radiation beam tube Stainless steel screw from the top grid plate Steel parts from the rotary specimen rack assembly Fe-55 Ni-63 Zn-65 Fe-55 Ni-63 Fe-55 Ni-63 Fe-55 Ni-63 not calculated 1,3E+02 1,9E+04 7,5E+05 8,8E+02 1,1E+05 3,1E+04 3,2E+04 1,0E+04 2,1E+02 8,3E+03 5,0E+02 2,5E+05 6,9E+04 1,8E+04 5,6E+03 7,2E+06 1,3E+07 3,2E+06 4,4E+02 3,6E+05 2,9E+09 3,3E+06 Calculated value/ measured value

12 When the results of the calculation are compared to the results of the analyses they are largely in concurrence for and Fe-55. The same goes for Ni-63 except for the values for the bottom part of the graphite blind element. The stainless steel screw and the central radiation beam tube are represented best by the AKAT calculation. The greatest uncertainty occurs when estimating the specific activity in the graphite blind element. CONCLUSIONS AND OUTLOOK As it is planned to dismantle the reactor facility completely by hand, it is necessary to have realistic radiological data in order to prepare for the dismantling procedure. Furthermore, both the release of radioactive materials into the environment and the costs for external disposal of the radioactive waste from the dismantling of the reactor are to be kept to a minimum. The results from taking the samples have already shown that it is possible to dismantle the reactor facility with the usual tools without having to use special remote handling techniques. Since the details of the neutron flow used in the AKAT calculations are based on old information, which only applies to the core area, it is planned to carry out new, more dimensional calculations about the distribution of the neutron flux. These calculations will take into consideration especially the components outside the reflector. Because the reactor is constructed symmetrically two dimensions suffice (deterministic neutron transport calculation). A three-dimensional model is necessary only in the area of the external radiation beam tube, possibly using the Monte Carlo procedure. Once again the composition of the material is to be determined in more detail and more data on the cross-sections are to be used for calculating the activation. The additional calculations aim to do the following: In order to keep the radiation danger to dismantling staff to a minimum in concordance with the new Radiation Protection Regulation the entire inventory of activity is to be known in the planning phase. During the dismantling phase as few radioactive materials as possible are to be released into the air and water of the surrounding area by minimizing the number of active and contaminated parts which need to be taken apart. In order to adhere to building regulations certain limits regarding measurements and masses must not be exceeded when moving the dismantled reactor components through the building. In order to do this, the activity inventory of the components to be moved must be known as well as possible. The activity inventory must be measured as exactly as possible outside the reflector on an axial level with the core. From this it can be determined in the planning phase of dismantling the reactor whether a partial dismantling of the reactor tank and the biological shield will be sufficient. How much of the reactor needs to be dismantled can be planned exactly and thus significantly reduce the amount of radioactive waste. This would make it possible to considerably reduce the costs for the dismantling of the reactor and interim storage of radioactive waste.

13 Fig. 1. Schematic view of the places where samples were taken

14 Fig. 2. Chipping a sample from the graphite blind element in the sample box Fig. 3. Concrete drill with closed water cycle during the radial drilling