Radioactive effluent releases from Rokkasho Reprocessing Plant (1) - Gaseous effluent -

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1 Radioactive effluent releases from Rokkasho Reprocessing Plant (1) - Gaseous effluent - K.Anzai, S.Keta, M.Kano *, N.Ishihara, T.Moriyama, Y.Okamura K.Ogaki, K.Noda a a Reprocessing Business Division, Japan Nuclear Fuel Limited Aza Okitsuke, Obuchi, Rokkasho-mura, Kamikita-gun, Aomori-ken, Japan. Abstract. In Japan, Rokkasho Reprocessing Plant (RRP) is going to start the operation in service as the first large-scale commercial reprocessing plant of spent fuels which has annual reprocessing quantity of 800tUpr in maximum. RRP started active test with spent fuels on the 31st March In the active test, we verify the performance of reprocessing, removal of radioactive nuclides from gaseous and liquid effluent, and so on. When spent fuel assemblies are sheared and dissolved, radioactive gaseous waste containing 85Kr, 3H and 129I is released to the atmosphere. In order to limit the public dose as low as reasonably achievable, RRP removes radioactive materials by the help of scrubbing, filtering, etc, and then releases gaseous effluent through a main stack that allow to make dispersion and dilution very efficient. For active test, concerning the radioactive gaseous effluent to be released into the environment, the target values of annual release quantity have been defined in our Safety Rules based on the estimated annual release quantity at the design stage of RRP. By monitoring the radioactive material in exhaust, RRP controls it not to exceed the target values in order to keep the public dose as low as reasonably achievable. RRP will reprocess 430 tupr spent fuel tupr (about 460 fuel assemblies for PWR, and about 1250 fuel assemblies for BWR) during active test. The amounts of radioactive gaseous waste during active test are evaluated to be less than the target values. In addition, public dose from external exposure, inhalation, and, ingestion of agricultural and livestock food, influenced by RRP during active test is evaluated low sufficiently. KEYWORDS: radioactive gaseous effluent, radioactive effluent, Reprocessing Plant, public dose 1. Introduction Rokkasho Reprocessing Plant (RRP) has been started to build since 1993, which is the first large size commercial reprocessing plant in Japan. RRP is designed to be able to reprocess 800 tupr of spent fuel provided from light-water reactor (BWR, PWR). RRP has been carried out: Water test since 2001, Chemical test since 2003, Uranium test since After those, RRP started Active test since 31st.Mar In this test, we reprocessed spent fuel actually, and confirmed the safety function and performance of equipments concerning treatment of plutonium and fission products. Concrete confirmation items are amount of radioactivity released to environment, separation performance of fission products, distribution performance of uranium and plutonium, treatment ability of liquid and solid waste, and so on. In this paper, we describe the amount, evaluation, and public impact concerning discharged radioactive gaseous waste during 1st.Apr.2006~ 31st.May The control of radioactive gaseous waste at RRP 2.1 The outline of control of radioactive gaseous waste RRP is designed to decrease the amount of radioactivity in gaseous waste as low as reasonably achievable. To be concrete, the gaseous waste from each process is decontaminated by scrubbing, filtering, etc. After that, the gas is discharged from main stack that has effect of diffusion and dilution (See Figure1). As a result of the design to decrease these released radioactive materials, the public dose around RRP is evaluated to msv/y caused by the gaseous and liquid waste from RRP when reprocessed 800 tupr of reference spent fuel[1]. JNFL define the release control target value in the safety * presenting author, masaki.kanou@jnfl.co.jp 1

2 regulations of RRP based on the expected release amount per a year which is used in a public dose evaluation. At release of radioactive gaseous waste, JNFL measure and control the radioactivity so as to be kept always less than the control target values. (See Table 1) Figure 1: removing equipments for radioactive gaseous wastes PROCESS shearing &dissolution mist filter HEPA filter iodine filter HEPA filter separation off gas scrubber HEPA filter iodine filter U purification and denitration off gas scrubber HEPA filter ma in stack Pu purification and U-Pu co-denitration off gas scrubber HEPA filter iodine filter vitrification off gas scrubber HEPA filter iodine filter Table 1: Target values for release control of the radioactive nuclides Radioactive nuclides in gaseous effluent (Bq/year) 85 Kr 3.3x H 1.9x C 5.2x I 1.1x I 1.7x10 10 Total alpha 3.3x10 8 Total beta (gamma) 9.4x The measurement method of radioactivity in gaseous waste At RRP, monitoring equipment is installed by the stacks to measure the radioactivity in radioactive gaseous waste (See Figure 2). Kr-85 contained in gaseous waste is measured continuously by gas monitor. The other nuclides in gaseous waste is collected by appropriate method (conform to chemical and physical form) and measured periodically (See Table 2). Figure 2: Monitoring equipment for main stack On-line monitor A Control room server On-line monitor B Sampling equipment A Sampling equipment B Monitoring panel Terminal Main stack On-line monitors Gas monitor Sampling equipment Dust Iodine Ruthenium Dust monitor Iodine monitor Tritium Carbon 2

3 Table 2: sampling and measurement radioactivity in gaseous effluent. Nuclide Sampling Measurement Kr-85(GAS) GAS monitor (plastic scintillation detector) H-3 Oxidation furnace, cooling trap Liquid Scintillation Counter C-14 Oxidation furnace, organic solvent trap Liquid Scintillation Counter I-129, I-131 Charcoal cartridge Gamma spectrometer Total alpha, total beta Paper filter Gas flow counter 3. Evaluation method 3.1 The amount of radioactivity released The active test of RRP consists of 5 steps and Step 4 ended on 13th February Step 5 is going to be done. In the active test, spent fuel is reprocessed with lower burn-up to higher gradually as shown in Table 3. We evaluated the amount of radioactivity released to environment from the start of the active test (31st. March 2006) to the end of step 4(13th February 2008), and the public impact during 1st April 2006 to 31st March (Those are according to reprocessing of spent fuel.) The released nuclides were classified as the nuclide expected to be removed in design such as 129 I, 131 I, and the nuclides which are not expected to be removed in design such as 85 Kr, 14 C. Table 3: The specification of the spent fuel reprocessed in the active test number of amount of fuel step spent fuel spent fuel type assemblies [tu pr ] 1 Burn-up [MWd/t] cooling time[y] PWR ,000~17, (31/3/2006~26/6/2006) PWR ,000~33,000 10~18 2 PWR ,000~36,000 10~20 (12/8/2006~6/12/2006) PWR ,000~36,000 8~15 BWR ,000~21, BWR ,000~36,000 8~19 (29/1/2007~26/4/2007) PWR ,000~47,000 8~21 4 PWR ,000~48,000 5~18 (31/8/2007~13/2/2008) BWR ,000~40,000 8~18 shearing started on 1st April The released amount of the nuclides expected to be removed in design RRP has reprocessed spent fuel with the specification as shown in Table 3 during the active test. On the other hand, the target values are decided based on the release radioactivity corresponding to reprocessing 800 tu Pr of standard spent fuel a year*. So, in order to confirm the removal performance for radioactive gaseous and liquid waste expected in design, we estimated the EMR (Estimated Maximum release Radioactivity) value from measured amount of radioactivity, which corresponds to reprocessing 800 tu pr of standard spent fuel a year, and compared with the target values. EMR = A m x A s / A i A m ; the whole amount of radioactivity released in the period of the active test steps 1 to 4 A s ; amount of radioactivity of 800 tu Pr of standard spent fuel* 3

4 A i ; amount of radioactivity in the spents fuel reprocessed in the period of the active test steps 1 to 4 calculated by ORIGEN2 code * standard spent fuel: spent fuel type of PWR, burn-up 45,000MWd/tU Pr, specific power 38MW/ tu Pr, cooling time 4 years, defined by Application for Designation of Reprocessing Business (ADRB) [2] The released amount of the nuclides which are not expected to be removed in design In the design of RRP, released radioactivity of the nuclides which are not expected to be removed were evaluated on the assumption that whole amount of radioactivity will be released to environment. So, in the active test, we estimated the amount of radioactivity in the reprocessed spent fuel by means of ORIGEN2 code, and compared it with measured amount of radioactivity released. 3.2 Evaluation method for public impact We evaluated the dose to the public around RRP by released radioactivity from April 2006 to March 2008, using the standard weather condition at RRP, and social environment (eating habits, etc). 4. The result of evaluation 4.1 The amount of radioactivity released Table 4 shows the measurement results of the amount of radioactivity released in gaseous waste during Apr.2006~Mar The amount of released radioactivity was less than discharge target value at each one year (Apr.2006~Mar.2007, Apr.2007~Mar.2008). Table 4: The amount of released radioactivity to the atomosphere activity [Bq/y] nuclide Apr. 2006~ Apr. 2007~ target values [Bq/y] Mar Mar Kr x x x10 17 H-3 6.0x x x10 15 C x x x10 13 I x x x10 10 I x x x10 10 Total alpha ND ND 3.3x10 8 Total beta (gamma) ND ND 9.4x10 10 ND: not detected The result of evaluation to the nuclides expected to be removed in design In the nuclide expected to be removed in design, tritium, iodine-129 and iodine-131 were detected in the exhaust and the effluence in the period of the active test steps 1 to 4. Therefore, we evaluated the EMRs and compared between EMRs and the target values. As a result of comparison, EMR values of these nuclides were all less than each target value. (refer to Table 4) Concerning total alpha activity and total beta (gamma) activity, the measured values were less than detection limit in the period of the active test steps 1 to 4. The detailed result of evaluation is as follows. 4

5 [tritium (H-3)] 1.5x10 13 Bq of tritium (it is about 0.4% of total amount contained in spent fuel) was released to the atmosphere, and EMR was about 0.4% of the target value. In design of RRP, it is evaluated that about 10 % of total amount of tritium contained in spent fuel will be released to the atmosphere [3]. But EMR was much less than the target value. This difference might be due to the following two reasons. 1. There is a report that says the half amount of tritium in spent fuel will be transferred to hulls [4]. 2. The actual decontamination factor of tritium is better than expected value in design. [I-129] Whole HALW (High Activity Liquid Waste) which produced in the period of active test step 4 has not been vitrified yet. So, the amount of radioactivity of iodine-129 from vitrification facility which calculated in ADRB [2] is added to EMR of iodine-129 in order to estimate conservatively. 5.5x10 8 Bq of iodine-129 (it is about 1.4% of total amount contained in spent fuel) was released in the period of the active test steps 1 to 4, and EMR was about 50% of target value. In design of RRP, it is evaluated that almost all the amount of iodine-129 contained in spent fuel will be transferred to gaseous waste and removed by iodine filter, and that about 1% of total amount of iodine-129 contained in spent fuel will be released to the atmosphere [3]. But EMR was less than the target value, because the actual removal performance was better than that expected in design. [I-131] 1.1x10 7 Bq of iodine-131 was released to the atmosphere in the period of the active test steps 1 to 4, and EMR was about 0.4% of target value. In design of RRP, it is evaluated that iodine-131 will be originated by spontaneous nuclear fissions of curium in HALW, and removed by iodine filter, and that about 10 % of total amount of iodine-131 will be released to the atmosphere. But EMR was less than the target value, because removal performance was better than the performance expected in design actually. Table 5: The evaluation result of EMRs to the nuclides expected to be removed in design nuclide calculated value measured value target values EMR [Bq/y] (A i ) [Bq] (A m )[Bq] [Bq/y] H-3 3.5x x x x10 15 I x x x10 9 *1 1.1x10 10 I *2 1.1x x10 7 *2 1.7x10 10 April 2006 ~ February 2008 (the active test steps 1 to 4) *1: Added the amount of activity from HALW Vitrification Facility (3.7x10 9 Bq/y evaluated in ADRB [2] ) to evaluate conservatively. *2: Iodine-131 is originated from spontaneous nuclear fissions of curium. Therefore, discharged amount of iodine-131 was estimated by stored amount of curium, and EMR was calculated The result of evaluation to the nuclides that isn t expected to be removed in design In design of RRP, it is assumed that all the amount of krypton-85, carbon-14 released to the atmosphere. The amount of these radioactive nuclides need to be estimated appropriately for a reprocessing plan, so it is important to collect data about released radioactivity of these nuclides. We compared the amount of radioactivity in the spent fuel which has been reprocessed in the period of the active test steps 1 to 4 (from April 2006 to February 2008) calculated by ORIGEN2 code with measured radioactivity released. (refer to Table 6) 5

6 [Kr-85] 6.3x10 16 Bq of krypton-85 was released to the atmosphere in the period of the active test steps 1 to 4. This amount is about 120% of ORIGEN2 calculated value. [C-14] 3.0x10 12 Bq of carbon-14 was released to the atmosphere in the period of the active test steps 1 to 4. This amount is about 16% of total amount contained in spent fuel calculated by ORIGEN2 code. When calculation of total amount of carbon-14 contained in spent fuel, we defined the content of nitrogen in UO 2 pellet as 50ppm. (This value was used for setting target value per year in ADRB [2].) On the other hand, actual content of nitrogen in UO 2 pellet of fuel reprocessed in the period of the active test steps 1 to 4 was about 10 ppm. So, the amount of activity in spent fuel close to actual amount of activity released by re-calculation by setting the content of nitrogen as 10 ppm. However, content of nitrogen in UO 2 pellet will not be constant, so it is conservative to make a reprocessing plan based on expected released amount of carbon-14 calculated by content of nitrogen as 50ppm. We intend to acquire data and to evaluate it continuously in future, too. Table 6: The evaluation result of ratio between measured value and calculated value to the nuclides which are not expected to be removed in design nuclide Calculated values by measured values (B) ORIGEN2 (A) [Bq] [Bq] ratio B/A Kr x x C x x st March 2006 ~ 13th February 2008 (the active test steps 1 to 4) 4.2 Public impact The dose to the public in these periods was evaluated to 5.4x10-4 msv/y (during April 2006 to March 2007) and 1.3x10-3 msv/y (during April 2007 to March 2008). Both of the dose are sufficiently less than 1 msv/y that is dose limit for public. Table 7 The evaluation result of public dose caused by radioactive gaseous and liquid waste Dose to the Public [msv/y] Apr. 2006~Mar Apr. 2007~Mar Gaseous effluent 4.9x x10-3 Liquid effluent 4.6x x10-4 Total 5.4x x Conclusion RRP was confirmed to have satisfactory performance to remove radioactive materials expected in design and to be able to reprocess 800 tu pr /y with satisfaction in the target values, that is, with keeping the public dose as low as reasonably achievable. REFERENCES [1] K.Sasaki, S.Tsuura, S.Murakami (Japan Nuclear Fuel Limited), Full papers of 10th International congress of the International Radiation Protection Association, " Assessment of Dose caused by Releases of Gaseous and Liquid Waste from Rokkasho Reprocessing Plant in Normal Operation ",

7 [2] Japan Nuclear Fuel Limited, " Application for Designation of Reprocessing Business ", 1989 (revised in 1996, 2001, and 2004) (in Japanese) [3] Japan Nuclear Fuel Limited, etc., "The behavior of radioactive nuclide at reprocessing plant", JNFS R Revision 1, 1996: ADRB literature at evaluation of appendix 7 (in Japanese) [4] Tanehiko Yamanouchi (Japan Nuclear Cycle Development Institute), etc. "The behavior of tritium at reprocessing plant", TN , 1981 (in Japanese) 7