Atomic Energy, Vol. 94, No. 3, 2003
INVESTIGATION OF 85Kr EMISSION AND ITS EFFECT ON THE RADIATION CONDITIONS DURING THE PREPARATION OF BOR-60 FUEL FOR RECOVERY
V. V. Serebryakov, A. P. Kirillovich, A. F. Sviridov, and I. Yu. Zhemkov
UDC 621.039.58:621.039.516.4
The main stages of the BOR-60 closed fuel cycle, implemented on the experimental base at the ScientificResearch Institute of Nuclear Reactors, are examined. The 85Kr emission at the stages of preparation of the spent BOR-60 fuel assemblies for recovery is determined experimentally. It is shown that the maximum 85 Kr emission as a result of destruction of fuel element cladding with oxide uranium fuel is 68%; its contribution to the irradiation dose to the public as a result of mechanical disassembly of the fuel elements in a single BOR-60 fuel assembly with 10% burnup and a 10-yr holding time does not exceed 1·10–4% of the dose limit (1 mSv/yr).
Investigations aimed at the development and improvement of the processes involved in the secondary use of recovered nuclear fuel, obtained by dry technology and fabrication of fuel elements from it by vibrational compaction, in the fuel cycle have been ongoing for a long time at the Scientific-Research Institute of Nuclear Reactors [1, 2]. The unique experimental and research base makes it possible to solve problems associated not only with developing technological processes for recovering irradiated fuel but also solving safety problems. The urgency of these problems is underscored by the strategy for the advancement of nuclear power, where the importance of gradually increasing the fraction of renewable energy sources in the fuel–energy balance of the country is emphasized. Closed Fuel Cycle. The first demonstration experiment on reprocessing a large batch of irradiated uranium and uranium–plutonium fuel in a closed fuel cycle (Fig. 1), including preparation of fuel for recovery, pyroelectrochemical process for reprocessing the spent fuel, preparation of experimental fuel elements and fuel assemblies for reuse, was conducted at the Scientific-Research Institute of Nuclear Reactors in 2000–2002. Investigations aimed at obtaining the experimental data required for ensuring radiation safety of the processes occurring in the closed nuclear fuel cycle were conducted during these stages. In the present paper, we present the results of an investigation of the escape and behavior of gaseous fission products, specifically, 85Kr, at the first stages of the process, which consist of preparing fuel for recovery. The gas emissions at the stage of mechanical preparation of fuel elements, which is the most informative stage with respect to events, is examined in detail. The preparation of fuel for recovery included three stages (Fig. 2). The first stage is preparation of spent fuel assemblies. This stage was conducted using equipment placed in a protective chamber of the process section, where a bundle of fuel elements was freed, using remote controlled mechanical methods, from the structural components of the fuel assembly: casings, lattices, and cribs. After the fuel assemblies were disassembled, the fuel elements were fed into the process section of the materials-engineering investigations, where the lateral screens and gas collectors from the active part, followed by mechanical chopping (crushing), into 10–15 mm long fragments, were separated by remote control in a protective chamber using special equipment. The final stage in the preparation of fuel for recovery is oxidation, which separates the fuel composition from the steel cases by oxidative recrystallization at 400–600°C. Federal State Unitary Enterprise, Scientific-Research Institute of Nuclear Reactors. Translated from Atomnaya Énergiya, Vol. 94, No. 3, pp. 220–226, March, 2003. 1063-4258/03/9403-0179$25.00 ©2003 Plenum Publishing Corporation
179
Fig. 1. The scheme of a closed fuel cycle using the experimental base of the Scientific-Research Institute of Nuclear Reactors: 1) BOR-60; 2, 3) spent fuel assemblies to holding and disassembly, respectively; 4) fuel elements to preparation; 5) fuel in fuel-element fragments to recovery and fuel assembly fabrication; 6) fuel assemblies with recovered fuel in a reactor; 7) storage of high-level wastes; 8) solid and liquid radioactive wastes; 9) fresh fuel.
Fig. 2. Scheme of the technological process for preparing fuel for recovery: 1) spent fuel assembly; 2) preparation of the fuel assembly; 3) preparation fuel elements; 4) oxidative recrystallization; 5) oxidized fuel to recovery; 6) terminal parts of the fuel assemblies to solid wastes; 7) mixture of gases and aerosols; 8) catching of aerosols on filters in the protective chambers; 9) mixture of gases and aerosols in the ventilation center.
Highly active structural material, process wastes, and cladding fragments were sent to conditioning or final storage, depending on the content of fissioning materials. The gas-aerosol emissions removed on ventilation systems of the objects were subjected to two-step purification (first step – filter of the protective chamber, second step – filtrational station of the ventilation center at the Scientific-Research Institute of Nuclear Reactors). Characteristics of Spent BOR-60 Fuel Assemblies. Four spent fuel assemblies were selected for recovery: two with pelleted fuel consisting of highly enriched uranium dioxide – up to 90% 235U (type U) and two with vibrationally compacted mixed oxide uranium–plutonium fuel – plutonium fraction up to 19% and 235U enrichment up to 71% (type UP). The 180
TABLE 1. Main Computed Characteristics of BOR-60 Reactor Spent Fuel Assemblies Activity, GBq Fuel assembly
U-1
Burnup, %
10.5
Holding time, yr
85
fission products
actinides
total
Kr
23.7
6.10·104
1.22·101
6.01·104
9.55·102
5
2.48·10
1
1.04·10
5
2.58·103
U-2
11.9
10.4
1.04·10
UP-1
11.6
12.5
1.04·105
8.65·104
1.90·105
2.25·103
11.6
5
4
5
2.28·103
UP-2
11.5
1.05·10
9.67·10
2.02·10
fuel assemblies contain 37 6-mm in diameter fuel elements with 0.3 mm thick cladding. The fuel element construction has two gas collectors: bottom and top collectors, 400 and 500 mm long, respectively. The neutron-physical characteristics of the fuel assemblies were calculated using the TRIGEX computer code based on the BNAB-90 constants library and the CONSYST-2 system for preparing the constants, the isotopic composition and the radiation characteristics of the fuel were calculated using the AFPA computer code [3, 4]. The error in determining the actinide mass is 8–16% and the error for fission products mass is 20–25% (Table 1). The results of the radiochemical analysis of 235U content in reprocessed fuel confirm the accuracy of the computed values. The discrepancy was less than 3%. This gives grounds for assuming that the computed values used in the work are reliable. On the basis of the computational investigations for all reprocessed fuel assemblies, the contribution of five radionu90 clides ( Sr, 90Y, 137Cs, 137mBa, and 147Pm) to the activity of fission products is 91–97%. The maximum computed activity is 1.05·105 GBq per fuel assembly. The radioactivity of gaseous fission products is determined primarily by 85Kr (on the average, about 2 TBq per fuel assembly) and, to a lesser extent, by 129I and 3H. It is well known that tritium in fast reactors diffuses through the fuel element cladding into the coolant and the gas cushion, and its residual activity in the fuel elements is less than 0.0185 GBq per 1 kg of fuel [5]. The 129I activity in uranium–plutonium irradiated fuel after a holding period of 10–12 yr is estimated to be 2.9·10–3 GBq, and the amount in the gas phase of the fuel element reaches 3.7·10–5 GBq per 1 kg of fuel [5]. Investigation Procedure. Since the basis of the activity of gaseous fission products of irradiated fuel assemblies is 85 Kr (more than 99%), whose β decay is accompanied, primarily, by electron emission, its escape in the technological operations was monitored by pumping gas through the flow-through ionization chamber (BDBG-02P, sensitive volume 10 liters). The measurement principle is based on counting individual β particles followed by measurement of the ionization current produced by these particles [6]. As a result of the low quantum yield accompanying the decay of 85Kr (<0.45% per decay), it is difficult to use other detection blocks, based on detection of photon radiation, because of the high cost and the difficulty of arranging the blocks at the sampling site. The BDGB-02P detection block, which is a component in automated radiation-monitoring systems, which operate on objects also in the ventilation center at the Scientific-Research Institute of Nuclear Reactors, provides a continuous regime for measuring the activity of radioactive gases in emissions from ventilation systems. A calibration conversion factor 1.3–10–6 m3/(Bq·sec), calculated for the spectral distribution of electrons emitted as a result of 85Kr decay [6], was used to take account of the energy dependence of the detection efficiency of β particles in calculations of the activity of radioactive gases. The absolute values of the emissions were calculated taking account of the background stationary emission levels and the required corrections (concentration, conversion, and so on factors). The admissable main measurement error for the volume activity of 85Kr was ±20% [6]. Gas Emission at the Fuel-Assembly Disassembly Stage. According to the data from the radiation monitoring system, one case of short-time emission of gaseous fission products was detected during the removal of the casing and disassembly of four fuel assemblies. Comparing the detected event and the cyclogram of the process established that gas entry occurred during the disassembly of the casing of the UP-1 fuel assembly as a result of damage to individual fuel elements, located in the cutting zone. When the fuel elements were cut open, radioactive gas escaped into the protective chamber and then was removed into the ventillation system. The detected activity of 85Kr entry through the ventillation pipe into the atmosphere was 181
Fig. 3. Variation of the volume activity of gaseous inputs into the ventillation center during the mechanical preparation of batches of irradiated fuel elements.
98 GBq, which is 4% of the computed activity of the gas in a fuel assembly. The contribution of this stage to the total emission of radioactive gases over the entire period of disassembly of four fuel assemblies (months) did not exceed 0.1%. Gas Emission During Mechanical Disassembly of Fuel Elements. The results of continuous monitoring were used to study the behavior of 85Kr, beginning with escape from the open fuel elements into the protective chamber and up to emissions from the ventillation center. Figure 3 shows as an example a graphical illustration of the indications of the automated system of radiation monitoring of gaseous emission; the figure shows the change in the volume activity of the radioactive gas during mechanical disassembly of nine irradiated fuel elements (UP-type fuel assemblies). The change in the volume activity makes it possible to detect clearly the moment when the fuel element cladding becomes unsealed. The duration of the growth and dropoff of the indications, equal on the average to about 20 min, is comparable to single exchange of air in the volume of the protected chamber. The maximum and minimum peaks are explained by a special feature of the technological process, performed in a certain order, in accordance with which the lower gas collector is separated first and then the upper gas collector is separated. In Fig. 3, the peak corresponding to separation of the top gas collector is singled out in the form of an enlarged fragment. Integrating the separate intervals of the peak changes in the volume activity in time and comparing the results with the cyclogram of the process, we obtained the computationalexperimental estimates of the krypton emission during the process of opening up fuel elements (Table 2). The results show that irrespective of the fuel type the 85Kr emission during opening up of the fuel element cladding shows close values and equals on the average 61%. This agrees with [7], where for uranium fuel the krypton emission resulting from penetration of the cladding was 69.6%. The remaining gas can be released during oxidation and recovery. The total activity of 85Kr emission over the entire period of disassembly of four fuel assemblies was 4689 GBq. The procedure used for preparaing fuel assemblies prevented the emissions of radioactive gases from the object from undergoing excursions above the daily levels. The contribution to the total emission of radioactive gases through the ventillation pipe was 1.5%. Additional measures to catch radioactive krypton will need to be taken to increase the volume of fuel elements to be prepared. Analysis of the data obtained by monitoring gaseous emissions showed that the time over which the gas medium from the protective chamber, where the casing of the fuel assemblies was removed and the fuel elements were opened, reaches the tall stack is about 2 min. This period of time must be taken into account when developing protective measures which ensure that an additional purification system is connected up quickly and events occurring at objects at the Institute can be identified on the basis of their classification. 182
TABLE 2. 85Kr Emission at the Stage of Separation of the Active Parts of Fuel Elements from the Screens Fuel assembly Parameter U-1
U-2
UP-1
UP-2
955
2580
2250
2280
Detected total activity, GBq
629
1460
1200
1400
Emission per fuel assembly, %
67.9
56.6
53.3
61.5
17
39.5
32.4
37.9
Computed activity, GBq
Activity per fuel element, GBq: average maximum
28.5
51.4
40.7
40.2
minimum
11.2
27.8
26.7
27.5
Fig. 4. Activity of expected 85Kr emission at the stage of mechanical separation of fuel elements from 1 U-1 fuel assembly of a BOR-60 reactor as a function of fuel burnup and holding time: 1 yr (1), 6 yr (2), and 12 yr (3).
Gas Emission During Fuel Oxidation. Gas emissions during oxidation were not detected. This is due to the incomplete emission of gas from the fuel matrix at process temperature 400–600°C. This is confirmed by experimental data from [7], where it is shown that the emission of the gas phase (85Kr) accompanying heating of the uranium fuel up to 1650°C was about 22%. Since the accumulation of fission products is associated with fuel burnup and the half-life of 85Kr (10.7 yr) is longer than the irradiation time of fuel assemblies in the reactor (about 2.5 yr), the computed and experimental data on the fuel assemblies investigated were used to construct a linear plot of the accumulation of 85Kr in fuel assemblies and the emission of 85Kr as a function of the fuel burnup and holding time. However, these approximate relations give values which are too high compared with the actual values, because the calculations neglected the times between reactor runs, when the activity of fission products decreases as a result of radioactive decay. Figure 4 shows as an example a family of curves, characterizing the activity of 85Kr emission at the stage of mechanical preparation of fuel elements in one fuel assembly from a BOR-60 reactor with different burnup and holding time, for which the 85Kr emission was 68%. These curves are convenient for practical calculations and estimates of the radiation safety of such operations. Assessment of Radiation Effects on the Public. The gaseous emissions into the atmosphere rise above the cutoff section of the stack and almost immediately start to mix with the air, become diluted, and migrate together with air masses 183
in the direction of motion of the air streams. We used the MPA-98 predictive computer code to assess the consequences of radioactive-gas emission into the atmosphere and the effect of the radioactive gas on the inhabitants in the region of the Institute [8]. It was established using this code that for short-time emission of 85Kr with activity 1460 GBq into the atmosphere (maximum total activity, detected at the stage where the fuel elements of the U-1 fuel assemblies were disassembled), the expected irradiation dose to the public, according to conservative estimates (with the worst meteorological conditions) does not exceed 1·10–6 mSv, which is 1·10–4% of the main dose limit for irradiation of the public (1 mSv). Conclusions. The main stages of the BOR-60 closed fuel cycle, implemented on the experimental base of the Scientific-Research Institute of Nuclear Reactors, were examined. The 85Kr emission at the stages of preparation of spent BOR-60 fuel assemblies for recovery was determined experimentally. It was shown that its maximum emission during destruction of the fuel element cladding with oxide uranium fuel was 68%. The unsealing of individual fuel elements during removal of the casing of the fuel assemblies also can result in emission of 85Kr gas into the volume containing the protective equipment. The calculations and the experimental data were used to construct a linear plot of the activity of the expected 85Kr emission at the stage of mechanical disassembly of fuel elements in a single fuel assembly with different values of the burnup and holding time. The 85Kr contribution to the irradiation dose to the public resulting from disassembly of a single BOR-60 fuel assembly with 10% burnup and 11-yr holding period does not exceed 1·10–4% of the dose limit (more mSv/yr). Analysis of the radiation conditions at the stages of preparation of irradiated fuel of a BOR-60 reactor for recovery shows that the technological processes, developed and used in practice and implemented using remote controlled experimental setups, ensure that the allowed levels and limits regulated by normative documents and requirements will not be exceeded. The results obtained can be used as reference data for substantiating the safety of a closed fuel cycle and in post-reactor studies of spent fuel assemblies and fuel elements.
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