Planning of Radiation Protection Precautionary Measures in Preparation for Dismantling and Removal of the TRIGA Reactor at the Medical University of Hannover Gabriele Hampel, Uwe Klaus. Department of Nuclear Medicine, Medical University of Hannover, Carl-Neuberg-Str., D-3065 Hannover, Germany E-mail: hampel.gabriele@mh-hannover.de. Fa. Babcock Noell Nuclear GmbH, D-97064Würzburg, Germany Abstract. At the Medical University of Hannover a research reactor of type TRIGA I was in operation between 973 and 996. The fuel is removed. It is now planned to dismantle the facility, remove the individual parts, take any radioactive waste to an external location and then to release the facility from the auspices of the German Atomic Law []. In order to carry out the decommissioning it is necessary to submit verification of adequate radiation protection precautionary measures which are based on the present radiological condition of the reactor facility, the dismantling techniques, the handling and amount of the expected radioactive waste. For the determination of the present radiological condition, samples were taken from the reactor components and both the activation and dose rates were calculated. A catalog of contamination and dose rates was drawn up. It is intended to use manual dismantling techniques and to minimize cutting of activated components. The radioactive waste will be collected, sorted, packed suitably, transported through the building, declared and then released for interim storage to an external location. For all of these steps the radiation protection measures will be described and the expected dose rates, the individual and collective doses of the staff, the radiation exposure of the not direct with the decommissioning project involved MHH personal and the general public will be estimated. In addition, the radiation exposure of small children and adults has been calculated at the most affected point where effects from spent air containing radioactive substances could be expected, under both normal and unusual operating conditions during the dismantling of the facility.. Introduction Since 973 the Clinic for Nuclear Medicine at the Medical University of Hannover (MHH) had operated a research reactor of type TRIGA I at a maximum power level of 50 kw mainly to produce radio-pharmaceuticals and for activation analyses. In 997 the facility was shut down. Two years later all of the spent TRIGA fuel elements were returned to the United States []. It is now planned to dismantle the facility, remove the individual parts, take any radioactive waste to an external location and then to release the facility from the auspices of the German Atomic Law []. In order to carry out the decommissioning it is necessary to submit verification of adequate radiation protection precautionary measures which bases on the present radiological condition of the reactor facility, the dismantling techniques, the handling and amount of the expected radioactive waste.. Present radiological condition of the reactor facility In order to determine the radiological condition of the facility after the irradiation, samples were taken from the reactor components [3] and both the activation and dose rates were calculated [4,5,6]. The main activated components are the core support, the graphite elements, the control rods, the graphite reflector with lower and upper grid plate, the rotary specimen rack, the radial beam tube, the central irradiation tube, the filter equipment, the reactor tank and the surrounding baryt concrete of the biological shield. The main data required for planning the dismantling of the reactor are the average and maximum specific activities in the irradiated reactor components, the dose rates at the surface and
at m distance from components and in the areas affected by dismantling and the dismantling thresholds in the reactor tank and biological shield as well. Samples were taken from the following materials and components from areas of high neutron flux (see Fig. ): Graphite from a graphite element Aluminium from the reactor tank Aluminium from the top grid plate Stainless steel from a screw in the top grid plate Baryt concrete and reinforcement irons from the biological shield 3 4 5 Fig.. Schematic diagram of the reactor. Samples were taken from the marked positions and materials: stainless steel from a screw in the upper grid plate, baryt concrete and reinforcement irons from the biological shield, 3 aluminium from the reactor tank, 4 aluminium from the upper grid plate, 5 graphite from a graphite element.
The samples were prepared and analysed at the VKTA Rossendorf Laboratory for Environmental and Radionuclide Analyses and at MHH using the following methods: High resolution gamma spectrometry with HP germanium detector (type n) Comprehensive beta measurement (low level alpha beta counter) Liquid scintillation spectrometry after radio-chemical separation for determining H-3, C-4, Fe-55 and Ni-63 Mass spectrometry with inductively coupled plasma (ICP-MS) for determining traces of chemical elements The calculations of the specific activities and dose rates were done by the IKE Stuttgart [4, 5,6]. The D S N program DORT and the 3D Monte Carlo program MCNP-4C were used to solve the neutron and photon transport equation. For calculating the activity after irradiation and determination of photon sources ORIGEN- was used. Cross sections processed at IKE were used based on ENDF/B- VI and JEF-. for transport calculations and FENDL for activation calculations. The principle geometry of the TRIGA reactor is cylindrical and therefore it can be described by a D model. The non-symmetrical parts, i.e. the radial beam tube, the filter equipment, graphite elements, control rods and instrumentation tubes as well require a 3D description. Table I shows the calculated specific and total activities of the important components and materials of the TRIGA reactor. Table I. Materials and activity of components Component Material Specific activity [Bq/g] Total activity [Bq] Upper grid plate AlMg3F8 3.9E+04.0E+08 Lower grid plate AlMg3F8 5.3E+04 4.E+08 Central irradiation tube AlMg3F8.8E+05 3.6E+08 Instrumentation tubes AlMg3F8 5.0E+03.0E+08 Reactor tank with radial AlMg3F8 9.0E+0.E+08 beam tube Rotary specimen rack Stainless steel, Al.5E+07 3.3E+0 Steel components Stainless steel.7e+07 3.4E+09 (screws etc.) Graphite elements Graphite, Al 9.3E+04 3.3E+09 Control rods B 4 C, Al.5E+05 4.6E+08 Filter equipment Al, Pb, He, graphite 4.E+03.4E+09 Graphite reflector Graphite, Al.0E+04.6E+0 Biological shield (D Baryt concrete 0.0 30 model) Biological shield (3D model) Baryt concrete. 7,700 The maximal total activity of the activated components and materials of the TRIGA reactor is about 84 GBq. This value includes the measured activity of the reinforcement irons of the biological shield. The main parts of the total activity are the rotary specimen rack and small stainless steel components, the graphite reflector and activated concrete of the biological shield. The stainless steel components have the highest specific activities, but a total mass of only.5 kg. The graphite reflector has less specific activity, but a mass of 800 kg and the activated concrete also has a low specific activity, but a heavy mass of 0,000 kg. 3
Fig. shows the activation in the reactor tank and the biological shield. Fig.. Top: Co-60 activation product in the aluminium tank with boundary for dismantling, Bottom: right: Ba-33 activation in the symmetric region of the biological shield with boundary for dismantling, left: Ba-33 activation product in the top layer of heavy concrete near the radial beam tube 4
From the specific activity the nuclide vectors in table II were determined. Table II. Nuclide vectors for the relevant materials Material Nuclide Portion [%] Aluminium Stainless steel Fe-55 Co-60 Ni-63 Eu-5 Fe-55 Co-60 Ni-63 Graphite H-3 C-4 Co-60 Eu-5 Eu-54 Baryt concrete H-3 Fe-55 Co-60 Ba-33 Eu-5 Eu-54 7 7 3 6 5 65 6 6 5 64 6 Table III shows examples of the results of the dose rate calculations. The calculated dose rates of a stainless steel screw as a function of the distance can be seen in Fig. 3. Table III. Dose rates at the surface and at a distance of meter for some components Material / Location Dose rate [msv/h] At the surface At a distance of meter Empty reactor tank, m over 5.0 E-03.8 E-03 bottom Lower grid plate 6 E+00 8 E-0 Graphite reflector 9.5 E+00 9 E-0 Rotary specimen rack E+0 4. E-0 Central irradiation tube 6 E-0 3.0 E-03 The calculations and measurements provide a complete picture of the activity of the area near the core from the center of the core to the biological shield. Thus all the main data required for planing the dismantling techniques of the reactor, handling the residual materials and the radiation protection measures are available. Additionally the amount of radioactive waste and its total activity can be estimated. 5
3. Basic safety goals The concept of MHH for decommissioning and dismantling of the reactor is based on the following goals: Use of common and manual dismantling techniques Minimisation of the cutting of activated components Reduction of the amount of radioactive waste by radiological measurements for release Minimisation of the radiation exposure for the working staff and the not involved persons (e.g. patients, medical personal) Prevention of pollution to the environment Maintenance of operational safety, fire and physical protection Maintenance of a safe dismantling under both normal and unusual operating conditions 4. Dismantling concept All radioactive materials of the reactor will be removed with the aim to release the facility from the German Atomic Law []. Due to the results of the measurements and calculations of the activity inventory the concept for dismantling the TRIGA reactor is based exclusively on the mechanical dismantling and taking apart of the reactor facility with manually used tools usual in the trade. Most of the build-in components can be dismantled in accordance with the regular operational procedures under water. It is planed to minimize the cutting of activated or contaminated components. For example the reflector will be attached by its shackles and lifted from the reactor tank into a special waste cask without any cutting. It is planed to dismantle the reactor in 7 phases. The schedule for decommissioning is shown in table IV. Table IV. Schedule for decommissioning No. Phase of dismantling Time in weeks Preparations 4 Built-in components of the reactor tank 8 3 Water cooling and purification system 4 4 Reactor tank and biological shield together with the 3 radial beam tube 5 Other systems like pneumatic transfer system 6 6 Final measures 4 7 Radiological measurements for the release of the 3 reactor facility and remaining structures Sum 5 The reactor tank, the biological shield and the radial beam tube will be dismantled only partial. The reactor tank will be cut into segments and separated from the biological shield with a severing tool. In order to avoid the spread of contamination, this work and the dismantling of the biological shield and the radial beam tube will be carried out in a foil tent with separate ventilation and aerosol filter. 6
5. Precautionary radiation protection measures The following measures are planned to ensure the radiation protection during dismantling: Shieldings (mobile and fix) Mobile ventilation systems Remote control techniques Foil tends Decontamination of components Overhead of areas and components with foils Personal protection cloth like overall, extra gloves and inhalation protection Packages for the radioactive residual materials and prevention of surface contamination of the packages In order to estimate the dose during dismantling procedure the calculated dose at the surface and at a distance of m of the components was taken as a basis to definite the following dose guidelines for the working staff: Daily 0. msv Weekly 0.5 msv Monthly.0 msv After the German Radiation Protection Ordinance [7] a total dose of 0 msv is the yearly limit of a working person. Addionally the following values of dose rates at the working positions are determined for radiation protection measures: Dose rate [µsv/h] Radiation protection measures <.5 Not necessary.5 to 00 Decision of the officer for radiation protection > 00 Necessary 6. Monitoring The monitoring includes measurements of the dose rates determination of the activity concentration in the air in working areas sampling of the components during dismantling measurements of the external radiation exposure of the personnel incorporation control of the personal control of the activity release with air and water surveillance of the environment The personnel s exposure to radiation will be solely from external radiation exposure. Incorporations during dismantling are not relevant since the release of aerosols is less and radiation protection measures are taken into account. The whole body counter of MHH is used to determine the incorporation of gamma radionuclides of the staff at the begin and at the end of their work. 7
7. Estimation of the radiation exposure The estimated collective dose for the working staff is 5 msv. 95 % of this value belongs to the 9 persons working for dismantling, radiation protection and handling the radioactive waste. For non involved persons in other areas of the radiological building outside the controlled area adjacent to the reactor facility there will be no significant radiation exposure. Based on German Radiation Protection Ordinance [7] the radiation exposure of persons in the surroundings and the area of the MHH was calculated using the administrative regulation (AVV) [8]. First the radiation exposure due to the emissions of airborne radioactivity from MHH facilities - not including the MHH TRIGA reactor - was calculated at selected points in the area and the surroundings of the MHH. Secondly, for the calculation of the radiation exposure caused by the dismantling of the TRIGA reactor, dose coefficients for several nuclides and selected points were determined. The results show that the effective dose for children ( year) due to the maximal emissions of airborne activity (Co-60 and H-3) from the TRIGA reactor in a year at the unfavourable point is less than µsv [9]. The regulatory limits for the general public of the German Radiation Protection Ordinance [7] will not be exceeded. 8. Summary The results show that the safe limits in accordance with the German Radiation Protection Ordinance [7] will not be exceeded for any persons as long as the usual radiation protection measures are adhered to. This, then, guarantees the prerequisite of radiation protection precautionary measures for carrying out the dismantling and removal of the TRIGA reactor. The radiation protection measures to be carried out during the dismantling and removal procedures will be stipulated in a Radiation Protection Instruction. Once all of these measures have been concluded, including the final measurements for releasing, the reactor facility can be released from the auspices of the German Atomic Law []. Reference // German Atomic Law from 5. July 985 // H. Harke, G. Hampel, U. Klaus, L. Lörcher, Radiation Protection during Handling the spent TRIGA Fuel at the Medical University of Hanover, WM 00 Conference, February 7 - March, 000,Tucson, AZ /3/ G. Hampel, H. Harke, U. Klaus, W. Kelm, L. Lörcher, Sampling and Radiological Analysis of Components of the TRIGA Reactor at the Medical University of Hannover, WM 0 Conference, February 5 - March, 00,Tucson, AZ /4/ G. Hampel, F. Scheller, W. Bernnat, G. Pfister, U. Klaus, E. Gerhards: Calculation of the Activity Inventory for the TRIGA Reactor at the Medical University of Hannover (MHH) in Preparation for Dismantling the Facility, WM 0 Conference, February 4 8, 00, Tucson, AZ /5/ W. Bernnat, E. Gerhards, G. Hampel, U. Klaus, G. Pfister: Two and Three Dimensional Activity and Dose Rate Calculations for the TRIGA Reactor at the Medical University of Hannover (MHH), Jahrestagung Kerntechnik, Stuttgart (Mai 00), S. 55-58, ISSN 070-907 /6/ W. Bernnat,, E. Gerhards, G. Hampel, U. Klaus, G. Pfister Measurements and Calculations for Determination the Activity Inventory and Dose rates of the TRIGA Reactor at the Medical University of Hanover, st World TRIGA Users Conference, June 00, Pavia, Italy /7/ German Radiation Protection Ordinance from July 0 th 00 /8/ Administration Regulation to 45 of the German Radiation Protection Ordinance from February st 990 /9/ G. Hampel, U. Klaus, W.-H. Knapp, W. Stratmann Radiation Exposure of the Public due to the Emission of airborne Radioactivity from the Medical University of Hannover, 34. Jahrestagung des Fachverbandes für Strahlenschutz 8