STUDY ON PARTITIONING OF LONG LIVED NUCLIDES FROM HLLW IN TSINGHUA UNIVERSITY

Transcription

1 STUDY ON PARTITIONING OF LONG LIVED NUCLIDES FROM HLLW IN TSINGHUA UNIVERSITY Chongli Song Institute of Nuclear Energy Technology, Tsinghua University, China Introduction The final disposal of radioactive waste is one of the key problems that effect on the further development of nuclear energy industry. High level liquid waste (HLLW) contains more than 95% of radioactivity in spent nuclear fuel elements. The current disposal technique in the world is to vitrify HLLW into a borosilicate glass matrix and to encapsulate the glass into a metal container. Quite large volume of highly active waste (HAW) would be then disposed into a deep repository to keep it apart from the biosphere. However, the 99 Tc, 129 I and transuranium (TRU) elements contained in HLLW have high toxicity and very long half-life. They dominate the long-term radiological hazard from a repository. The required geological containment time for the resulting waste might be many thousands or even millions of years. It brings about great uncertainties in long-term risk assessment. The vitrification of HLLW is a currently available industrial technique. However, because of the public acceptance, there is no deep repository that has been constructed until now. Partitioning and Transmutation (P-T) is a substitute technique for treatment and disposal of HLLW. The P-T concept [1] involves chemical separation of transuranium (TRU) elements as well as long-lived nuclides (for example, 99 Tc, 129 I, etc.) from HLLW, and transmutation of them to either stable or short-lived nuclides. The P-T concept constitutes an advanced nuclear fuel cycle. The P-T can not eliminate the need of the deep repository. However, the implementation of the P-T concept could significantly reduce the long-term risk of the radioactive waste disposed in the deep geological repository. It could also simplify the design and construction of the repository and allow the full use of the nuclear energy resources and of beneficial radioisotopes. The partitioning of TRU elements as well as 99 Tc from HLLW originating from the Purex process is one of the critical technical issues of the P-T concept. Several partitioning processes have been developed for separation TRU elements as well as longlived nuclides from HLLW, such as TRUEX process [2] in U.S.A., DiDPA process [3] in Japan, Diamex process [4] in French and TRPO [5] process in China. A clean use nuclear energy (CURE) concept was proposed for the back-end of the nuclear fuel cycle in recent year [6]. In the concept the partitioning requires not only to remove the TRU, 99 Tc and 129 I, but also to segregate 90 Sr and 137 Cs from HLLW. 89

2 After partitioning, HLLW would become a non-α, low and intermediate lever radioactive waste that would be suitable for shallow-land disposal. So in the CURE concept the required decontamination factor (D.F.) for TRU elements will be much higher than that in P-T concept. The partitioning of HLLW can be implemented with transmutation in the future to reduce the long-term risk of nuclear waste. It can also be used as a pretreatment method of HLLW to reduce α waste and highly active waste (HAW) volume. In this paper the partitioning used as a waste volume reduction method will be discussed and the study on the partitioning of TRU elements in Tsinghua University will be reviewed. Partitioning as a Waste Volume Reduction Method Required Decontamination Factor for Commercial HLLW Partitioning used as a pre-treatment for HLLW is possible. The objective of partitioning is to separate TRU elements, 99 Tc as well as 90 Sr and 137 Cs. After partitioning the original HLLW would be de-classified to a non-α, low and intermediate lever radioactive waste that would be suitable for shallow-land disposal. The required decontamination factor (D. F.) is given in Table 1 for a typical commercial spend nuclear fuel that had a burn-up of 33000MWd/tU, a cooling time of 10 years and 99.75% of U and Pu had been removed in reprocessing. In table 1 the α- waste standard for individual package ( Bq/kg) is chosen and 0.13m 3 glass (specific weight of 2) or 0.40m 3 concrete waste is supposed to be produces after solidification of the declassified HLLW. If the non α-waste standard is chosen the required D. F. for TRU elements should be reached to for vitrification and for cementation. In table 1 the D. F. value for 129 I is not included because during reprocessing almost all iodine was evaporated and would not be appear in HLLW. In order to get a higher waste volume reduction factor, the separation of lanthanides (Lns) and actinides are necessary. The required D. F. for TRU elements in Lns fraction should be higher than [7]. Table 1. The Required D. F. for Treatment of Typical Commercial HLLW to a Waste Suitable for Shallow Land Disposal Nuclides Specific Chinese standard GB American regulation 10CFR61 activity in Standard Required D. F. Standard Required D. F. HLLW Bq/kg Vitrification cementation Ci/kg Vitrification Cementation TRU Tc Sr Cs Volume Reduction Effect of Partitioning for Commercial HLLW Partitioning could get remarkable waste volume reduction if the required D. F. 90

3 can be fulfilled. For a typical commercial HLLW the salt content of main process streams after partitioning is listed in Table 2. It could been seen from the table that only 6.6% salts in HLLW would become α- waste or HAW and about 90% salts in HLLW could de-classified into MAW or LAW. The α-waste volume reduction factor would be higher than 30 and the HAW waste volume (including α waste) reduction factor would be quite higher [8]. Table 2. The Salt Composition of Main Process Streams of Typical Commercial HLLW after Partitioning (Spend Fuel Burn-up of 33000MWd/tU, Cooling Time 10 Years and 99.75% of U and Pu Removed by Purex Process, Specific Volume is 500l/tU and Salt Content is 250g/l) Waste α waste Non α HAW LAW or MAW after cooling Reuse Total Group Am + Cm + Np + 99 Tc Sr Cs RE Other F. P. Other salt U + Pu - F. P. Elemental weight, g/t U Salt content in HLLW, g/l ~90 ~ ~250 Salt percentage % ~ Volume Reduction Effect of Partitioning for Chinese Highly Saline HLLW Chinese highly saline HLLW is defense waste. The waste volume reduction effect of partitioning for Chinese highly saline HLLW had been discussed elsewhere [9]. The required D. F. for de-classifying the highly saline HLLW to a non-α, low and intermediate level waste is much low than that for commercial HLLW. The required D. F. is given in Table 3. The α- waste standard in Table 3 is for the individual package ( Bq/kg). Table 3. The Required DFs for Chinese Highly Saline HLLW [9] Nuclides Total α 99 Tc 90 Sr 137 Cs required DF {7} Table 4. Distribution of Salt of Chinese HLLW after Partitioning [10] Waste α waste Non α HAW LAW or MAW Reuse Total Group Am, Np, 99 Tc (including RE), 90 Sr Cs Other salts U and Pu Total salt content Salt content, g/l % The distribution of salt in the HLLW after partitioning is given in Table 4. Only 4.2% of salts originated in Chinese highly saline HLLW would be HAW that should be 91

4 vitrified and disposed in deep repository. More than 90% salts in the HLLW could be de-classified to MAW or LAW that could be solidified by cementation and disposed in shallow land with engineering bared. The pretreatment of Chinese highly saline HLLW by partitioning could get quite high HAW volume reduction factor and save the disposal cost. Partitioning Study in Tsinghua University The study on partitioning of HLLW in Tsinghua University began in Quite a lot of extractants were screened for its extraction ability for transuranium elements. A tri-alkyl phosphine oxide (TRPO) was found to be a very good extractant for tetra- and hexa-valent transuranium elements in nitric acid concentration from 0.05 to 7 M [10]. It was also an effective extractant for tri-valent plutonium, americium and curium at low and moderate HNO 3 concentration (<2mol/l). But the extraction of penta-valent NpO 2 was very poor. The extraction properties of TRPO for fission product elements were also studied. Among the F. P., Zr is highly extractable. The extraction ability of 30% TRPO for Tc(VII) is very high at low and medium nitric acid concentration, but it drops sharply with increase of nitric acid concentration. The extraction of Mo is sufficiently high and is similar to Tc(VII). Tri-alkyl phosphine oxide is a commercial available industrial extractant. Its trade name is TRPO in China and Cyanex 923 in Canada. TRPO contains C 6 -C 8 mixed alkyl and the average molecular weight is 346. Its density is g/cm 3, melting point is -21 and boiling point is 180 to 225 at 2-4mm Hg. Its solubility in water is low than 0.1g/l. However, It is miscible with common aliphatic hydrocarbons in all proportion. Until a total dose of Gy was absorbed, no significant change in extraction properties occurred [11]. It is also confirmed that there is no any back extraction difficulty for Np, Pu, Am and Cm in keeping the TRPO loaded with TRU elements for a week in the batch hot test with real HLLW [12]. The TRPO process for Removal of TRU Elements from HLLW A TRPO process was developed in Tsinghua University for removing TRU element from commercial HLLW [13, 14]. At beginning, amino-carboxylic complex TTHA-lactic acid was used as stripping agents for back-extraction of loaded TRU elements [13]. The cross contamination of TRU elements between the stripping streams was higher than 10%. In 1980s the extraction behavior of key element-americium was studied and a mathematical model was developed [15]. The model was used to calculate and to optimize the process parameter for extraction and stripping of americium. The stripping of TRU elements were also studied and the TRPO process flowsheet was modified [16]. In the flowsheet, 5.5M HNO 3 was used to strip Am (Cm) together with Lns, 0.6M H 2 C 2 O 4 was then used to strip Np and Pu and final uranium was back extracted by 5% Na 2 CO 3. The TRPO process flowsheet was verified with simulated HLLW skipped with TRU elements [5]. The hot tests of the TRPO process were carried out with HLLW of WAK at Institute for Transuranium Elements (ITU) Joint Research 92

5 Center of European Union at Karlsruhe Germany [17]. The hot test was completed with 24 stages of miniature centrifugal contactor in hot cell. The miniature centrifugal contactor had a rotor diameter of 10 mm with a hold-up about 4 ml each stage. The residence time for one stage was less than 1 minute. The flowsheet and some of parameters for first run of hot test were given in Figure 1 [17]. In the first run test the feed acidity was 0.7 M HNO 3 and the extraction section for TRU elements was 6 stages. Very good D. F. results were obtained for TRU elements except neptunium. For it the D. F. value was only In the second run of hot test, the feed acidity was increased to 1.36 M HNO 3 and 10 stages were used for the extraction of TRU elements. The D. F. value of 4100 for Np was obtained. The results for other TRU elements were also very good (see Table 5). TRPO 30%(v/v) 50ml/h HLLW 5000l/tU 0.7M HNO ml/h Scrub 1 1.0M HNO 3 169ml/h Stripping 1 5.5M HNO 3 44 ml/h Raffinate 115 ml/h Loaded TRPO 50 ml/h Scrub 2 0.1M HNO 3 13 ml/h Stripping2 0.6M H 2C 2O ml/h Am+Cm 40 ml/h Stripping3 5% Na 2CO ml/h Loaded TRPO 50 ml/h Waste 13 ml/h Np+Pu 50.6ml/h U 53.2ml/h TRPO used to recycling 50 ml/h Fig. 1. The flowsheet of the first TRPO hot test [17] Table 5. The Decontamination Factors of TRU elements, 99 Tc and Nd in the TRPO Hot Tests HNO 3 in feed Extraction stages Decontamination Factors Np-237 U-238 Pu-239 Am/Pu-241 Am-243 Tc-99 Nd-144 Run > 5400 > 760 > 2800 > 900 > 1400 > Run > 4100 > 7000 > 950 > 3200 > 760 >1700 > The Total Partition (TP) Process for Partitioning of Chinese Highly Saline HLLW In 1990s a total partition (TP) process was developed in Tsinghua University for 93

6 pretreatment of Chinese highly saline HLLW [9]. Its salt content is as high as 370g/l. The purpose of the pretreatment of Chinese HLLW is to declassify the HLLW to low and media level liquid waste that could then be cementation and shallow land disposal according Chinese regulation. It means that not only TRU elements and long-lived 99 Tc should be separated but also 90 Sr and 137 Cs need to be separated from the HLLW. Overcome the difficulty caused by high salt content, the TRPO was successfully used in TP process to remove TRU elements and 99 Tc from the Chinese HLLW [18]. A crown ether (dicyclohexano-18-crown-6) strontium extraction (CESE) process [19] was developed to separate 90 Sr and a potassium titanium hexacyanoferrate (II) (KTiFC) [20] ion exchanger was used to segregating 137 Cs. The flowsheet of the TP process is given in Fig. 2 [21]. The feed solution was 2.7 times diluted highly saline HLLW and its acidity was regulated to 1 M HNO 3. Fig. 2 The General Flow Sheet of Total Partitioning (TP) Process [21] A mathematical model was developed for the americium extraction in TRPO process [22] and for strontium extraction in CESE process [23]. The model was used to calculate and to optimize the TP process parameters. The TP process was verified with simulated HLLW skipped by TRU elements as well as with genuine HLLW. The hot test of the TP process was carried out in a compacted centrifugal contactor set and a small ion exchanger column in hot cell of Tsinghua University. The set consisted of 50 stages of miniature centrifugal contactors with a rotor size of 10-mm including 12 extraction stages for TRU and 10 extraction stages for Sr. Very good results were obtained in the hot test [24]. The D. F. of TRU was to 588 (see Table 6) and that for Sr and Cs was >2500 and >200 respectively. It means that after partitioning the HLLW became a non-α intermediate level liquid waste and it 94

7 could be solidified by cementation and disposed in shallow land with engineering barrier. Treatment of Chinese high saline HLLW by partitioning can get very high HAW waste volume reduction. After partitioning with TP process, about 7.8kg TRU oxides (including rear earth oxides) and about 10 kg waste oxide for cesium and strontium would be produced for 1 m 3 of HLLW. If they are solidified by vitrification about 39kg TRU waste and 59 kg HAW for Sr and Cs would be produced. However if the HLLW was directly vitrified, about 1380 kg of vitrified waste would be produced. The HAW waste volume reduction factor would be 14. Table 6. The DFs Obtained in the Hot Test of TP [24] Nuclides Total α 241 Am 239 Pu 237 Np U 99 Tc 90 Sr 137 Cs Material balance, % DF >2500 >200 The auxiliary processes of the TP process are now studied in Tsinghua University. They include the denitration and calcination process for Am (RE) stripping solution, Np/Pu separation process, the conversion process for uranium stripping solution and immobilization process for Cs-loaded KTiFC ion exchanger. The advanced extraction equipment such as pulsed column and centrifugal contactor for the TP process are also being studied. The separation of lanthanides and actinides The separation of trivalent lanthanides (Lns) and actinides is a difficult subject in the separation chemistry. However the separation of Lns and Actinides is necessary no matter how for the P-T concept or the clean use of nuclear energy concept. The study on the separation chemistry of lanthanides and actinides is one of research subjects of INET. In 1990s the extraction behavior of Am and fission product Lns were studied with varied extractants. Cyanex 301 had been proven to be an effective extractant for the separation of trivalent Am from Lns [25]. The separation factor between Am and Eu in the presence of macro amount of Ln was 500 provided the Cyanex 301 was saponified. The main composition of Cyanex 301 is bis (2,4,4 trimethylpentyl) dithiophospninic acid (HBTMPDTP). The content of R 2 PS 2 H is 75-83%, R 3 PS 5-8%, R 2 PSOH 3-6% and un-know compound ~2%. Some impurity in commercial Cyanex 301 can extract both Am(III) and Lns(III) with strong bonding at low ph. It would effect on the separation of Am and Lns in tracer amount of Lns. A >99% purified HBTMPDTP (>99%) could be obtained by purification of Cyanex 301 [26]. The HBTMPDTP is an S-coordinated extractant and prefers to extract Am rather than Ln. The extraction chemistry of Am and Lns was studied. The separation factor increases to 5000 for trace amount of Am and Eu when HBTMPDTP is used as an extractant. An empirical model of distribution ratio for Am and Lns was derived and an equations for calculating hydrogen ion concentration in aqueous phase was deduced. A 95

8 computer program for counter current separation of Am/Lns by HBTMPDTP extraction was compiled [7,27]. It can be used to calculate the process parameter and concentration profile in each stage. The program was verified with batch multistage counter current extraction experiments. A conceptual Am/Lns separation flowsheet by HBTMPDTP extraction was proposed [28] for the Am/Lns fraction from partition process of HLLW (see Fig. 3). Calculate results indicated that the more than % was extracted by HBTMPDTP while only 0.14% Lns was extracted. The D.F. for Am from Lns was that could meet the separation requirement. Fig 3. The conceptual flowsheet for Am/Lns separation by HBTMPDTP extraction The feasibility of the separation flowsheet was verified with a cross flow extraction hot test [27,29]. The feed solution was the Am/Lns fraction obtained from the hot test of total partition process for Chinese HLLW. After denitration and removing impurity, the feed solution was adjusted to ph 3.5. More than % of Am was extracted into the organic phase with 4 stages of cross extraction. The Am concentration in the 96

9 raffinate was 1 Bq/mL. Only ~3% Lns was extracted by HBTMPDTP. The average separation factor between Am and Lns was 3500 for each stage. The hot test results proved that the separation process was effective. The synergic extraction and separation of Am and Lns by HBTMPDTP/TBP- Kerosene was also studied. At ph about 2.8, quite high separation factor for Am/Lns could be obtained. A multistage counter current cascade experiment was performed. It included 7 extraction stages, 3 scrubbing stages and 2 stripping stages. Americium was effectively separated from Lns. The separation factor of Am from Lns was and the separation factor of Lns from Am was 2500 [30]. The integrated process for reprocessing and partitioning Nowadays reprocessing of spent nuclear fuel requires obtaining high recovery and high purity for uranium and plutonium. So the reprocessing process is very complicate and expensive. The integration of reprocessing and partitioning process would simplify the back end of the nuclear fuel cycle and would make it more reasonable and economical. The integrated process based on the Purex process for nuclear fuel reprocessing and the Chinese TRPO partition process is studying in Tsinghua University [31,32]. Six cases including direct connection of the Purex process and TRPO process and mixed TBP-TRPO process were reviewed. The computation programs for the Purex process and the TRPO process were established. The review showed that integrated process was a promising process and needed further investigation. Key data for the extraction behavior of TBP-TRPO-kerosene were determined experimentally. The results showed that the mixed TBP-TRPO extractant can effectively separate TRU elements as well as Tc from feed solution of spent nuclear fuel element. References 1. Croff, A. C.; Blomeke, J. O. Actinide Partitioning-Transmutation Program, Final Report, ORNL 5566(1980) 2. Horwitz, E. P.; Kalina, D. G.; Diamond, H.; Vandegrift, G.; Schulz, W. W. Solvent Extr. Ion Exch. 1985, 3, Kubota, M.; Nakamura, H.; Tachimori, S.; Abe, T. and Amono, H. "Removal of Transuranium Elements from High-Level Liquid Waste" IAEA-SM-246/24, (1981) 4. Madic, C.; Blanc, P.; Condamines, N.; Baron, P.; Berthon, L.; Nicol, C.; Pozo, C. et al "Actinide Partitioning from High-Level liquid Waste Using DIAMEX Process" The fourth International Conference on Nuclear Fuel Reprocessing and Waste Management, Recod'94: Proceedings vol. 3, London (UK), April 1994, CEA-conf Zhu, Yongjun and Song, Chongli "Recovery of Neptunium, Plutonium and Americium from Highly Active Waste, Tri-alkyl phosphine Oxide Extraction" Chapter 32 in Transuranium Elements: A Half Century, Edited by L.R. Morss and J. Fuger, ACS, Washington D.C. USA, (1992). 97

10 6. S. E. BINNEY, et al "CURE: Clean Use of Reactor Energy" WHC-EP-0206 Westinghouse Hanford Company, Richland WA 99352, Zhu, Y.; Chen, J.; Jiao_R. "Hot Test and Process Parameter Calculation of Purified Cyanex 301 Extraction for Separating Am and Fission Product Lanthanides" Global 97, Vol.1, , Yokohama, Japan, October 5-10, Song Chongli, "Partitioning of HLLW-An effective method for highly active waste volume reduction" to be published. 9. Song, Chongli, The concept flowsheet of partitioning process for the Chinese high-level liquid waste Atomic Energy Science and Technology, 1995, 29, (in Chinese) 10. Jiao. R.; Wang, S.; Fan, S.; Liu, B.; Zhu, Y.; Zheng, H.; Zhou, S.; Chen, S. Chinese J. Nucl. Radiochem. 1985, (in Chinese). 11. Xing, R., Song, C. to be published. 12. Song, C., Glatz, J-P., He, X., Bokelund, H., Koch, L., "Actinide Partitioning by means of the TRPO Process", RECORD'94, April 1994, London UK 13. Zheng, H.; Zhou, S.; Chen, S.; Jiao, R.; Wang, S.; Liu, B.; Fan, S.; Zhu, Y. Chinese J. Nucl. Sci. Eng. 1985, 5, 147(in Chinese). 14. Song, C., Zhu, Y., Yang, D., He, L., Xu, J., Chinese J. Nucl. Sci. Eng., 1992, 12 (3), 225 (in Chinese). 15. Song C L, Glatz J-P " Mathematical Model for the Extraction of Americium from HLLW by 30% TRPO and its Experimental Verification" A Value Adding Through Solvent Extraction, v2 The University of Melbourne, Australia, Song, C, Xu, J., Zhu, Y. Chinese J. Nucl. Radiochem. 1992, 14, 193 (in Chinese). 17. Glatz, J-P., Song, C.; Koch, L.; Bokelund, H.; He, H. Hot tests of the TRPO Process for the Removal of TRU Elements from HLLW Global'95, v1, 548, Versailles, France Sept Song, C.; Jin, G.; Wang, J.; Zhu, Y. Tsinghua Science and Technology, 1996, 1, He, L.; Weng X.; Yang, D. Chinese J Nucl. Eng., 1995, 15 (3): 76 (in Chinese). 20. Jiang, C.; Wang, S.; Song, C. Chinese J. Nucl. Radiochem, 1995, 17 (2): 99(in Chinese) 21. Song, C.; Wang J.; Liang, J. Treatment of High Saline HLLW by Total Partitioning Process Global 97, Vol.1, , Yokohama, Japan, October 5-10, Chen, J.; Wang, J.; Song, C. Nucl. Scien and Tech. (China) 1996, 7(3), 1 (in English) 23. Wang, Q.; Wang, J.; Song, C.; Tang, W. Chinese J Nuclear Radiochemistry, 1996, 18(2), 89 (in Chinese). 24. Song, C.; Wang, J.; Jiao, R. "Hot test of total partitioning process for the treatment of high saline HLLW, Global 99 Proceedings of the International Conference on Future nuclear systems, Aug.29-Sept.3, 1999 Jachson Hole, USA. 25. Zhu, Y. Radiochimica Acta. 68, 97, Chen, J.; Jiao, R.; Zhu, Y. Chinese J. Applied Chem. 1996, 13(2), 46 (in Chinese). 98

11 27. Chen, J.; Zhu, Y.; Jiao, R. Separation of Am(III) from fission product lanthanides by bis(2,4,4-trimethyl pentyl)dithiophosphinic acid extraction ---process parameters calculation, Nuclear Technology, 1998, 122(1): Chen, J., Jiao, R., Zhu, Y. "A conceptual flowsheet for Am/Lns separation by HBTMPDTP extraction " to be published. 29. Chen Jing Jiao Rongzhou Zhu Yongjun_A cross-flow hot test for separating Am from fission product lanthanides by bis(2,4,4-trimethylpenthyl) dithiophosphinic acid_radiochimica Acta, 1997, 76, Wang, X., Zhu, Y.; Jiao, R. to be published 31. Zhu, Y.; Han, B.; Wu, Q. "Study on the integration of the Purex process and the TRPO process in nuclear fuel cycle backend" Global 99 Proceedings of the International Conference on Future nuclear systems, Aug.29-Sept.3, 1999.Jachson Hole, USA 32. Wu, Q.; Han, B.; "Research on integration process of nuclear fuel cycle backend" 13 # Radiochemical conference, April 1998, Czech Republic 99