Waste retrieval of historical institutional radioactive waste from near surface repository

Transcription

1 Waste retrieval of historical institutional radioactive waste from near surface repository László Juhász a*, Péter Ormai b, Sándor Kapitány b, István Barnabás b, Károly Bérci c a National Research Institute for Radiobiology and Radiohygiene, Anna u. 5, H-1221 Budapest, Hungary. b Public Agency for Radioactive Waste Management, Puskas Tivadar u. 11, H-2040 Budaors, Hungary. c ETV-Erıterv, Angyal u. 1-3, H-1094 Budapest, Hungary. Abstract. The Hungarian Radioactive Waste Treatment and Disposal Facility (RWTDF) for disposal of institutional radioactive waste has been operating since The facility accommodated the radioactive waste arisen in the country between as well. The near surface repository is considered to be unsuitable for some of the waste types (e.g. long-lived spent radioactive sources) previously emplaced in the facility. Safety assessments have shown that these wastes may jeopardize the post closure safety; therefore corrective actions are needed before closure of the facility. After completing feasibility studies, decision was made to launch a safety enhancement programme. At first a trial project was to be performed with an aim to demonstrate the feasibility of the safety enhancement procedures. During the project, content of four concrete vaults of 70 m 3 each were to be processed. The planned measures include retrieval of waste from the disposal vault, sorting the safety critical items, putting them into interim store after conditioning or repackage. Before re-disposal of the waste packages that are free from long lived nuclides, volume reduction, repackaging and characterization should be carried out. Comparison the doses enduring by workers in the project with the avoidable doses presumed for the public in the future is the most important aspect in the assessment of the corrective actions. KEYWORDS: Radiation protection, historical waste, safety enhancement, radioactive waste management. 1. Introduction The radioactive waste disposal facility at Püspökszilágy is currently the only site for disposal of radioactive waste in Hungary run by the Public Agency for Radioactive Waste Management (PURAM). The repository was sited in 1971, designed and commissioned in 1976, according to the international guidelines in effect at that time. The repository is a typical near-surface facility, composed of concrete vaults and shallow wells for spent sealed sources. The disposal unit consists of 60 vaults each of 70 m 3 and six vaults each of 140 m 3. The total capacity of the repository is 5040 m 3. Initially, both unconditioned and conditioned wastes packaged in plastic bags or metal drums were placed in the disposal vaults and partly or fully grouted in-situ. Later the practice of grouting was terminated. In the absence of waste acceptance criteria, the repository has accepted almost all kinds of radioactive wastes generated during the utilisation of nuclear technology and isotope applications. By the end of 2004 the free capacity of the repository ran short. After the disposal vaults became full, the storage rooms in the treatment building that has recently been converted into a centralised interim store for institutional radioactive waste will be used until freeing space for the disposal of additional waste in vaults. * Presenting author, juhasz@osski.hu 1

2 2. Safety analysis Because the original licence did not deal with waste acceptance criteria (WAC) high activity sources and disused sealed radioactive sources (DSRS) consisting of long-half life and alpha emitting radionuclides have also been disposed of. The lack of defined WAC means that an acceptable benchmark was not established against which the type of waste received could be judged to be in conformance or non-conformance with a required standard other than external dose rate [1]. The safety of the facility previously has not yet been the subject of any comprehensive assessment that is why lately several safety assessments were carried out. The results of the safety assessments clearly indicate that the DSRS could result in high doses to individuals who intrude into the facility after the closure and they could also lead to high doses following any future disruption of the facility by natural processes. The presence of certain large sources (e.g. 137 Cs) in the vaults also gives cause for concern [2-3]. In the case of exposure resulting from past practices, the international recommendations (ICRP, IAEA) call for obligatory intervention above 100 msv/a and a more optimised (efforts and compliance) intervention where doses of between 10 and 100 msv/a are observed [4-5]. The basis for optimisation is the real dose associated with intervention activities vs. reduction of the potential dose in the future. Such an optimisation has never before been performed in Hungary. 3. Strategy for the corrective action PURAM started to plan a programme of intervention activities. Due to the rather large number of parameters involved, an optimised intervention programme had to be established on the basis of a system study [6-7]. The selected strategy can be summarized as follows: Recover wastes from vaults containing less than 10 m 3 of concrete backfill. Remove materials of safety significance and store on site pending disposal elsewhere. Condition other materials as necessary, including the application of low-force compaction where appropriate, and return to the vaults. Further institutional wastes appropriate for disposal in a near-surface facility are buffer-stored on the site pending appropriate conditioning and disposal using the space created by conditioning recovered wastes. After filling, vaults are backfilled with cement, ensuring that all space between and above the waste packages is filled. The vaults which are already backfilled are subjected to any remedial action needed to ensure that the backfilling for these vaults meets the same standard as the newly backfilled vaults. The waste recovery is to increase the site s long term safety by removing the vast majority of the wastes containing long half-life isotopes and emplacing them into an interim storage. The volume of the removed wastes is going to be decreased by volume reduction technologies. The repository capacity enhanced by this way is used to emplace the wastes to be accepted in the future. Reconditioning and repacking of the recovered waste may improve the local retention capability, may act as a chemical barrier and may provide possibility for volume reduction. In the case of the wastes that were not backfilled with cement, the retrieval is relatively easy, as compared with the backfilled wastes where safe retrieval of the spent sealed radiation sources would be considerably more onerous and risky. The corrective action would require the opening of the vaults that are already temporarily sealed and covered with protective layers of concrete, bitumen, clay and grass. 4. Demonstration program The main aim of the achievement of the demonstration program is that each step of the recovery technologies could be appropriately tested before starting the full scale recovery program. So cost, 2

3 volume of the disposal of the repackaged waste and timing are to be concretely defined. On the basis of the results of the demonstration program the final procedure and requirements of the retrieval may be set up [8]. Having received the necessary licenses for execution of the demonstration program, and finishing of the preparation works, the opening of the first vault took place on 16 th April, The task of the demonstration program is to retrieve the waste from four vaults designated for this program. These vaults are neighbouring (Figure 1.), and one of them (A13) is partially backfilled and the others are not backfilled. The calculated total activity of the four vaults is in the order of Bq, and mainly tritium and radiocarbon give this level of activity inventory. So the control measurements of these radionuclides are to be especially respected. Figure 1: The vaults (A11-A14) designated for retrieval in the demonstration program 1: inner tent 2: locking system 3: package delivery 4: outer tent 5: ventilation room 6: equipment room In the first step the cover soil and bitumen layer and later concrete layer and concrete panel roof were removed. Parallel to these works a double tent was built up above the vaults. During removing cover layers and before opening the vaults, the following measurements were carried out: - activity concentration of tritium in the moisture of the cover soil; - dose rate map above the vaults before removing concrete layer and after removing it; - gas sampling from vaults for defining the composition rate of tritium and radiocarbon; - activity concentration of tritium in the tent and at the reference point out of the tent. According to the results of the measurements the activity concentration of tritium ranged from 32 to 826 Bq/l in the humidity of the air of the vaults. The dose rate map is shown in the Figure 2. 3

4 The task has been started with two vaults where the waste was packaged in plastic bags (except for 3 drums), and in the third vault the waste was in different forms: plastic bags, drums, whole pieces of contaminated materials (metal, concrete, equipment, etc), and in the fourth vault the waste (mainly plastic bags) was grouted with cement. Until now the waste of the first three vaults has been recovered, and the waste from the fourth vault is under recovering. From the first two vaults the waste was removed by hoisting so each bag had been put in a drum. The contamination and dose rate of each drum were measured, and the date of recovering and mass of it were also indicated on the drum. Under the double tent only the hoisting of waste and the first characterisation was carried out. All other operations were conducted in the well equipped technology building. Figure 2: Dose rate map of the vault (A11) in µsv/h before removing of concrete layer After transportation of the drums to the technological building, firstly gamma spectrometry measurement and sampling for determination of the radionuclides of 3 H, 14 C, 90 Sr by using LSC have been performed on each waste bag. Then the waste was put in the glove box where it was separated as follows: - radiation source ( 60 Co, 90 Sr, 137 Cs, 226 Ra, 3 H target, etc.); - compactable waste; - non compactable waste; - special waste (glass, metal, liquid, ) According to the measurements the waste is put into separated drums when the waste contains the long half-life radionuclides (Ra, Th, U, radiocarbon), or tritium. The compactable waste is compacted into drums with low forces compactor (pressing force of 50 t), the non compactable waste is emplaced into a drum or a container regarding the size of waste, and then a qualification with waste assay system is used for the specification of the gamma emitting radionuclides and for determination of their activity. The total activity of the first two vaults was in the order of 10 TBq including all types of radionuclides. During a second separation the radiation sources have been separated regarding the activity and type (alpha emitting, tritium, etc.). Finally, the new packaged waste is to be qualified against the newly set WAC. 4

5 5. Radiation protection Concerning the high level activities of radiation sources (in the range of several 100 GBq) and much elevated radiation fields the demonstration program has to be concerned from the radiation protection point of view in both planning and performance. During the licensing, the received dose in proportion to the averted dose that should be acceptable, had to be verified. The most important radiation protection provision: every step must be controlled, so the involved workers, workplaces and equipment are supervised by a special dosimetry service. The values of dose rate around the vaults are very change depending on the type of the waste and the actual process, but the dose rate is in the order of few µsv/h. In the sorting premises the radiation field is much inhomogeneous regarding the quantity and the activity level of the waste moving in and out this workplace. In the most of time the dose rate is in the order of few µsv/h except the vicinity of glove box where the dose rate was in the order of 10 µsv/h owing to the drums in which the different waste has been collected. The contamination of drums, floor of workplaces and equipment is measured very frequently. Serious or durable contamination has not happened to date. The reference dose limits for the person involving in the demonstration program is 200 µsv per day and 400 µsv per week concerning the external exposure (evaluation is performed on the basis of the measurement of electronic dosimeter). In the case of external exposure control the personnel dosimetry comprises the compulsory (authority) film badge and the electronic dosimeter, as well as in the case of internal exposure control the examination of whole body and urine excretion for tritium have been used. For the retrieval of the first vault (A11) the planned collective dose was 12,7 man-msv and the actual value was 5,7 man-msv for 52 workdays. According to the authority film badge one result of the external exposure is shown in Table 1. The values of whole body measurements can be seen in the Table 2 and the results of excretion examination are summarized in the Table 3. Up to now the dose limit has not been exceeded [9]. Table 1: Received dose in the period of the retrieval of the first and second vault Results of the authority film badge in msv/2 months Wearing time Person No. 5-6/ / / / ,58 < 0,2 0,5 0, ,29 < 0,2 0,54 < 0,2 3. 0,29 0,37 0,72 0,3 4. 0,95 0,36 0,68 0,44 5. < 0,2 0,44 0,45 0,29 5

6 Table 2: The whole body examination after finishing of waste removing at the first and second vault (A11-12) The whole body measurements in Bq Measuring date Measuring date Person No. Cs-137 Co-60 Person No. Cs-137 Co-60 Ra-226 Th < 170 < < 160 < ±120 < < 130 < 140 < 330 < < 130 < < 140 < 100 < 250 < ±65 < ±170 < 140 < 310 < < 155 < < 170 < ±120 < 340 Table 3: Results of examination for tritium after finishing of the retrieval of the second vault (A12) Person No. Date of excretion Activity concentration of 3 H [Bq/dm 3 ] H E [µsv] ,8 0, ,3 0, , , ,3 0,38 6. Summary of the first experiences By analysing of the external and internal exposure data measured during recovery and processing of the waste it can be concluded that the results are rather favourable, the doses received by the workers were less than the values preliminary estimated, and the personal doses of the employees did not reach the time proportional value of the limit set for the radiation workers. The dosimetry service succeeded in blocking the spread of contamination, in shielding of high radiation fields and in pushing down the incorporation to low level The waste recovery operations have been executed under a double tent, from where the air may be released after filtering. During the other operations carried out in the technological building releases could happen only in a controlled manner. Discharges of the substances containing radioactive materials (air, water) increased only slightly but were much below the authorized limit and remained in the order of magnitude of the normal operation; hence the public was not affected by excess radiation. The radiation protection and environment monitoring systems were continually operating during the safety enhancement programme and did not register any significant difference as compared to the normal operational data. 6

7 The demonstration program will be completed by the end of followed that, the experiences gained during recovery and treatment of the waste, the impacts of the measures taken on the safety of the repository, the efficiency of providing free disposal capacity and the cost implications should be evaluated. On the basis of that, a decision can be made about which other vaults worth opening, and whether any modifications are needed in the technologies used, as well as which method is the most practical to redisposal of the segregated and reconditioned waste. REFERENCES [1] P. ORMAI, Safety Upgrading of the Püspökszilágy Disposal Facility, International Conference on Issues and Trends in Radioactive Waste Management, Vienna, 9-13 December 2002, Vienna (2003). [2] Safety Analysis of the Püspökszilágy Radioactive Waste Treatment and Disposal Facility: An Assessment of Post-closure Safety, Püspökszilágy Final Report, European Commission, PHARE , PH4.12/95(01)N2 (2001). [3] Safety Analysis of the Püspökszilágy Radioactive Waste Treatment and Disposal Facility, Püspökszilágy Final Report, ETV-ERİTERV, (2002) (in Hungarian) [4] INTERNATIONAL ATOMIC ENERGY AGENCY, Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management, INFCIRC/546, IAEA, Vienna (1997). [5] INTERNATIONAL COMMISSION ON RADIOLOGICAL PROTECTION, Radiation Protection Recommendations as Applied to the Disposal of Long - lived Solid Radioactive Waste, Publication 81, Pergamon Press, Oxford and New York (1998). [6] A HARPER, P DIXON, T GREEN, M KELLY, P SUMNER, K BERCI, L JUHASZ, F TAKATS A BAKSAY, Detailed study of waste retrieval and disposal options at the Püspökszilàgy Radioactive Waste Treatment and Disposal Facility, Serco Assurance and RWE NUKEM Consortium, PHARE project final report (2004). [7] P. ORMAI, Planning of safety upgrading measures at the püspökszilágy near surface repository, International Conference on the Safety of Radioactive Waste Disposal, Tokyo, Japan, 3-7 October (2005). [8] P. ORMAI, I. BARNABÁS, S. KAPITÁNY, First experience gained during the waste retrieval program at Püspökszilágy near surface repository, 33th Annual Meeting on Radiation Protection, Hajdúszoboszló (2008) (in Hungarian). [9] L. JUHÁSZ, L. BALLAY, O. TURÁK, P. ZAGYVAI, SZ. OSVÁTH, S. KAPITÁNY, P. ORMAI, Radiation protection experiences in retrieving of radwaste at Püspökszilágy, 33th Annual Meeting on Radiation Protection, Hajdúszoboszló (2008) (in Hungarian). 7